Text

Enclosures 1 and 2: Proposed Alternative Inspection Requirements for Steam Generator Nozzle to Reactor Coolant Pump Casing Welds and Alternative for Use of Code Case 648-2 for Inservice Inspection
	ML22308A158
Person / Time
Site:	Vogtle
Issue date:	11/04/2022
From:	; Southern Nuclear Operating Co
To:	; Office of Nuclear Reactor Regulation
Shared Package
ML22308A156	List: ML22308A157; ND-22-0817, Enclosures 1 and 2: Proposed Alternative Inspection Requirements for Steam Generator Nozzle to Reactor Coolant Pump Casing Welds and Alternative for Use of Code Case 648-2 for Inservice Inspection;
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	ND-22-0817
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Southern Nuclear Operating Company ND-22-0817 Enclosure 1 Vogtle Electric Generating Plant (VEGP) Units 3 & 4 Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1):

Alternative Inspection Requirements for Steam Generator Nozzle to Reactor Coolant Pump Casing Welds (VEGP 3&4-ISI1-ALT-17)

(This Enclosure consists of 38 pages, including this cover page)

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1. ASME Code Component(s) Affected Code Class: 1

Reference:

IWB-2500, Table IWB-2500-1 10CFR50.55a(b )(2)(xlii)

Examination Category: B-F Item Number: B5 .71

Description:

Alternative Inspection Requirements for Steam Generator Nozzle to Reactor Coolant Pump Casing Welds Component Number: Unit 3 Unit 4 SV3-SGA-Nozzle A-201 -96A SV4-SGA-Nozzle A-201-96A SV3-SGA-Nozzle B-201-96B SV4-SGA-Nozzle B-201-96B SV3-SGB-Nozzle A-201-96A SV4-SGB-Nozzle A-201-96A SV3-SGB-Nozzle B-201-96B SV4-SGB-Nozzle B-201-96B Drawing Number: Figures 1 and 2

2. Applicable Code Edition The First Interval of the Vogtle Electric Generating Plant (VEGP), Units 3 and 4 Inservice Inspection (ISI) Program is based on the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (BPV) Code,Section XI, 2017 Edition.

3. Applicable Code Requirement Subarticle IWB-2500 requires components specified in Table IWB-2500-1 to be examined. Table IWB-2500-1 requires a volumetric and surface examination of all NPS 4 (DN 100) or larger nozzle-to-component butt welds each inspection interval (Examination Category B-F, Item Number B5 .71). The applicable examination volume is shown in Figure IWB-2500-8(f).

In accordance with the provisions of 10CFR50.55a(b)(2)(xlii),Section XI condition:

Steam Generator Nozzle-to-Component welds and Reactor Vessel Nozzle-to-Component welds:

Licensees applying the provisions of Table IWB- 2500- 1, Examination Category B- F, Pressure Retaining Dissimilar Metal Welds in Vessel Nozzles ... Item B5.71 (NPS 4 or Larger Nozzle-to-Component Butt Welds) of the 201 la Addenda through the latest edition and addenda incorporated by reference in paragraph (a)(l)(ii) of 10CFR50.55a must also meet the following conditions:

(A) Ultrasonic examination procedures, equipment, and personnel shall be qualified by performance demonstration in accordance with Mandatory Appendix VIII.

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(B) When applying the examination requirements of Figure IWB-2500-8, the volumetric examination volume shall be extended to include 100 percent of the weld volume, except as provided in paragraph (b)(2)(xlii)(B)(l) of 10CFR50.55a:

(1) If the examination volume that can be obtained by performance demonstration qualified procedures is less than 100 percent of the weld volume, the licensee may ultrasonically examine the qualified volume and perform a flaw evaluation of the largest hypothetical crack that could exist in the volume not qualified for ultrasonic examination, subject to prior NRC authorization in accordance with paragraph (z) of 10CFR50.55a.

4. Reason for Request

In accordance with 10CFR50.55a(z)(l), relief is requested on the basis that the proposed alternative will provide an acceptable level of quality and safety.

Southern Nuclear Operating Company, Inc. (SNC) is requesting an alternative from the conditions listed in 10CFR50.55a(b)(2)(xlii)(B) to extend the examination volume to include 100 percent of the weld volume, and to seek NRC approval to use exception 10CFR50.55a(b)(2)(xlii)(B)(l) to ultrasonically examine the qualified volume shown in the 2017 Edition of Section XI, and to perform a flaw evaluation of the largest hypothetical crack that could exist in the volume not qualified for ultrasonic examination.

5. Proposed Alternative and Basis for Use

The API 000 design is unique in that the reactor coolant pump inlet nozzle is welded directly to the steam generator cold leg outlet nozzle (two per steam generator). The dissimilar metal circumferential butt weld joining the low alloy steel with austenitic stainless steel cladding steam generator nozzle to the cast austenitic stainless steel reactor coolant pump casing is classified as an ASME Section XI Class 1 weld (see Figure 1).

The API000 design has two steam generators and four reactor coolant pumps in each unit atVEGP.

SNC proposes to perform an inservice inspection encoded volumetric examination of the required 2017 Edition of ASME Section XI inspection volume, not the entire weld volume . The volumetric examination will be performed from the inner diameter (ID) surface. SNC will also perform the required surface examination on the outer diameter (OD) surface. The ultrasonic testing techniques will be qualified in accordance with the Performance Demonstration Initiative (PDI) Program which satisfies the requirements of ASME Section XI, Appendix VIII, Supplement 10, including 10CFR50.55a.

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In addition to the ASME Section XI examinations, SNC proposes to perform an eddy current examination from the ID surface. Although not an ASME mandatory examination, the eddy current examination utilized on the ID surface will be qualified in accordance with ASME Section V, Article 14 (2017 Edition).

This approach minimizes the impact of sound beam re-direction and scattering. The capability of detecting and length sizing of ID-initiated flaws in the weld and in the cast austenitic stainless steel (CASS) material was demonstrated on a representative blind test specimen. Therefore, full volume of the inner third of the weld, as required by the 2017 Edition of ASME Section XI, will be achieved. For clarification, SNC's proposed volumetric coverage for the First ISI Interval is depicted in Figure IWB-2500-8(f) and applied to the VEGP steam generator nozzle to reactor coolant pump casing butt welds in Figure 2.

The figures in the ASME Section XI, 2017 Edition (Figure IWB-2500-8 (c) - (f)) are illustrative with respect to the weld joint configuration. These figures are intended only to define the examination volume and examination surface extent for similar and dissimilar metal welds in components, nozzles, and piping. It is noted that the examination volume and examination surface extent is defined with respect to the weld (or weld end buttering) edges at the widest part of the weld (and weld end buttering) regardless of whether it is located on the inside or outside surface. For the examination volume, these weld (or weld end buttering) edges are extended to the inside surface and the 1/4-inch of adjacent base material is added to both edges to obtain the width extent of the examination volume. The 1/3t examination volume depth is taken from the inside surface. This approach ensures that the entire weld and weld end buttering width is captured in the examination volume regardless of the weld preparation configuration.

Figure 2 shows the proposed exam volume and that the widest part of the weld (and weld end buttering) is the same on the inside and outside surfaces. The weld is defined by the edges of the weld end buttering on the Reactor Coolant Pump Casing and the Steam Generator Nozzle. The 1/4-inch of adjacent pump casing and nozzle base material is taken from these weld end buttering points and the 1/3t depth is taken from the inside surface.

Figure 2 is to scale and the proposed exam volume captures all of the innermost weld. As noted in Figure 2, the entire weld and weld end buttering width is captured in the examination volume.

The examination procedure to be utilized has been qualified on a mock-up representative of the thickness and configuration of the steam generator outlet nozzle to reactor coolant pump casing weld and contains ID-initiated planar flaws in the weld, buttering and in the cast stainless steel material. Detection and length sizing qualification was extended to the full thickness. Because the examinations are performed from the ID surface, coverage of the examination volume is not limited. It is important to note, the examination volume is the inner 1/3 of the thickness of the weld and includes the weld and 0.25-inch of adjacent base metal on both sides of the weld and buttering. The weld and buttering on the steam

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--Alternative Provides Acceptable Level of Quality and Safety-Revision 0 (Page 4 of 8) generator outlet nozzle are composed of Alloy 690 weld metals. Alloy 690 weld metals used in the APIO00 design are much more resistant to developing flaws and have significantly improved flaw growth tolerance as compared to Alloy 600 weld metals.

Examination of the outer 2/3 is not required unless performing depth sizing of a flaw indication.

The ultrasonic testing and eddy current examination are the same as those applied to Reactor Pressure Vessel (RPV) nozzle to safe end dissimilar metal welds from the ID surface by the inspection vendor, except for the addition oflarger and deeper focused ultrasonic testing transducers. These added transducers allow for through-wall coverage through the full thickness of the weld in the event flaw indications are detected within the inner 1/3 of the thickness of the weld and adjacent base material or the defined examination volume. The AP 1000 steam generator nozzle to pump casing dissimilar metal butt weld is thicker than the RPV nozzle to safe end dissimilar metal welds found in other pressurized water reactors.

To extend the PDI qualification to this greater thickness and to account for the specific weld configuration, an AP 1000 steam generator to pump casing weld specimen was designed and fabricated by the Electric Power Research Institute (EPRI) in accordance with the EPRI/PDI Program. This specimen serves as a blind test specimen necessary to qualify the ultrasonic testing procedure and the ultrasonic testing personnel. The ultrasonic testing procedure and personnel qualifications are conducted by PDI under the PDI ASME Section XI, Appendix VIII program.

The eddy current examination techniques are qualified internally by the inspection vendor in accordance with ASME Section V, Article 14, intermediate rigor, using test data obtained from an additional APl000 steam generator to pump casing butt weld specimen, containing ID surface breaking planar flaws.

This combination of the ID surface applied ultrasonic testing and eddy current examination, that have been qualified or demonstrated, will allow detection of primary water stress corrosion cracking, the failure mechanism identified for dissimilar metal welds in operational pressurized water reactors. It is also noted that the ultrasonic testing techniques are capable of detecting, and length sizing, embedded planar flaws throughout the 2017 Edition of ASME Section XI examination volume as demonstrated on an open AP 1000 steam generator to pump casing butt weld specimen.

The preservice inspection (PSI) Interval examinations requested in Alternative VEGP 3&4-PSI-ALT-05 were complimented by the required ASME BPV Code Section III radiography examinations and the design organization's required ASME Code Section V ultrasonic testing imposed during component fabrication. The ultrasonic testing included in-progress inspections of the buttering material on both the steam generator and reactor coolant pump materials from the end face, and post-weld inspections of the full volume of the weld using multiple angles, four directional angle beam techniques, from both the

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ID and OD surfaces. The post-weld ultrasonic testing results were evaluated against the ASME Code Section III and Section XI standards for acceptance. No relevant indications were identified during the fabrication process or PSI examinations.

Enclosure 1 provides a non-proprietary copy of Westinghouse document LTR-PAFM 59-NP, which contains additional information and analysis that concludes a postulated defect in the outer 2/3 of the weld would not exceed the allowable flaw size over the licensed lifetime of the plant based upon the ASME BPV Code Section XI flaw tolerance evaluation and the ASME BPV Section III design evaluation. The proprietary non-redacted copy of this document was submitted to the NRC as Enclosure 3 to SNC letter ND-16-2542 (ADAMS Accession Number ML16355A222).

Enclosure 2 provides a non-proprietary copy of Westinghouse document L TR-P AFM 6, which provides justification as to why the constant flaw aspect ratios of 2 and 6 for axial and circumferential flaws was used in Enclosure 1. Enclosure 2 also provides justification as to why only flaw depth was evaluated in Enclosure 1. The proprietary non-redacted copy of this document was submitted to the NRC as Enclosure 2 to SNC letter ND-17-0113 (ADAMS Accession Number MLl 7032A524).

The proposed examinations are in accordance with the 2017 Edition of ASME Section XI, as described above. Therefore, SNC concludes that the proposed examinations will provide an acceptable level of quality and safety.

6. Duration of Proposed Alternative

The proposed alternative is requested for the First ISI Interval for Vogtle Electric Generating Plant (VEGP), Units 3 and 4.

7. Precedents

Vogtle Electric Generating Plant, Units 3 and 4 Preservice Inspection Proposed Alternative VEGP3&4-PSI-ALT-05 was authorized by an NRC SE dated April 17, 2017 (i.e., NRC Accession Nos. ML17097A337 and ML17097A450).

8. References

1. ASME Boiler and Pressure Vessel Code,Section XI, Division 1, "Inservice Inspection of Nuclear Power Plant Components," 2017 Edition.

2. ASME Boiler and Pressure Vessel Code,Section III, Division 1, "Rules for Construction of Nuclear Facility Components," 1998 Edition through the 2000 Addenda.

3. ASME Boiler and Pressure Vessel Code,Section V, Division 1, "Nondestructive Examination," 2017 Edition.

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4. Vogtle Units 3 and 4, Updated Final Safety Analysis Report (UFSAR), Subsection 5.1.3.3 .

9. Enclosures Enclosure 1: Westinghouse LTR-PAFM-16-59-NP, NRC RAI Response Regarding Inspection of APl000 Vogtle Units 3 & 4 and V.C. Summer Units 2 & 3 Steam Generator to Reactor Coolant Pump Suction Nozzle Weld (Non-Proprietary) Rev. 1 Enclosure 2: Westinghouse LTR-PAFM-17-6, Rev. 0, NP-Enclosure 1 (Non-Proprietary) - Additional Information Regarding APl000 Vogtle Units 3

& 4 and V. C. Summer Units 2 & 3 Steam Generator to Reactor Coolant Pump Suction Nozzle Weld Flaw Evaluation

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Figure 1 Steam Generator Channel Head Stea m Generator Primary Outlet Nozzle (Low Alloy Steel Clad with Austeniti c Stainless Steel)

L. Category B-F Weld per ASME Section XI (Diss imilar Metal Weld)

Reactm Coo~ ntPump Casing (Cast Austenitic Stainless Steel)

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Figure 2 OUTER DIAMETER I

1/4in. * *

*1/4in.

/

.... ?------

/ ---- ---

I I I I I I / I I I / I Reactor Coolant Pump Casing I I / I Steam Generator Nozzle (Cast Austenitic Stainless Steel) I \ I I (Low Alloy Steel Clad with Austenitic Stainless Steel)

. 1r**rUr*. .

CI T

ID 1/3t i I I I

__. . _______ qI ' :I ____________________________ j, r - - f--

F Examination Volume Cross-Section INNER DIAMETER

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--Alternative Provides Acceptable Level of Quality and Safety-Revision 0 Enclosure 1 Westinghouse LTR-PAFM-16-59-NP, NRC RAI Response Regarding Inspection of APlO00 Vogtle Units 3 & 4 and V.C. Summer Units 2 & 3 Steam Generator to Reactor Coolant Pump Suction Nozzle Weld Rev. 1

SVP _SV0_004439 ND-22-0817 Enclosure I - VEGP 3&4-ISII-ALT-17 Attachment 2 Westinghouse Non-Proprietary Class 3 Page 1 of 22 Page 11 of 38 LTR-PAFM-16-59-NP Revision 1 NRC RAI Response Regarding Inspection of APJOOO Vogtle Units 3 & 4 and V. C. Summer Units 2 & 3 Steam Generator to Reactor Coolant Pump Suction Nozzle Weld November 2016 Author: Alexandria Carolan* , Piping Analysis and Fracture Mechanics Rick Rishel*, Wesdyne Verifier: Anees Udyawar*, Piping Analysis and Fracture Mechanics Stephan Sabo*, W esdyne Approved: Benjamin Leber*, Manager, Piping Analysis and Fracture Mechanics

Electronically approved records are authenticated in the electronic document management system.

Westinghouse

SVP_SV0_004439 ND-22-081 7 Enclosure I - VEGP 3&4-ISII-ALT-17 Attach ment 2 Westinghouse Non-Proprietary Class 3 Page 2 of 22 Page 12 of38 LTR-PAFM-16-59-NP Revision 1 Record of Revisions Rev. Date Revision Description 0 September Original Version 2016 1 November Incorporate fabrication-related inspection data on flaw sizes from post-weld Radiographic Testing 2016 (RT) and Ultrasonic Testing (UT) examinations from Vogtle Units 3 and 4 and V.C. Summer Units 2 and 3. Major changes in Revision 1 are identified by vertical bars in the left-hand margins of the document.

Trademark Note:

APIOOO is a trademark or registered trademark of Westinghouse Electric Company LLC, its affiliates and/or its subsidiaries in the United States of America and may be registered in other countries throughout the world . All rights reserved. Unauthorized use is strictly prohibited. Other names may be trademarks of their respective owners.

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SVP_SV0_004439 ND-22-0817 Enclosure l - VEGP 3&4-ISil-ALT-17 Westinghouse Non-Proprietary Class 3 Page 3 of 22 Page I 3 of 38 LTR-PAFM-16-59-NP Revision 1 FOREWORD This document contains Westinghouse Electric Company LLC proprietary information and data which has been identified by brackets. Coding (a.c.e) associated with the brackets sets forth the basis on which the information is considered proprietary. These codes are listed with their meanings in WCAP-7211 Revision 8 (September 2015), "Proprietary Information and Intellectual Property Management Policies and Procedures. "

The proprietary information and data contained in this report were obtained at considerable Westinghouse expense and its release could seriously affect our competitive position. This information is to be withheld from public disclosure in accordance with the Rules of Practice 10CFR2.390 and the information presented herein is to be safeguarded in accordance with 10CFR2.903. Withholding of this information does not adversely affect the public interest.

This information has been provided for your internal use only and should not be released to persons or organizations outside the Directorate of Regulation and the ACRS without the express written approval of Westinghouse Electric Company LLC. Should it become necessary to release this information to such persons as part of the review procedure, please contact Westinghouse Electric Company LLC, which will make the necessary arrangements required to protect the Company 's proprietary interests.

The proprietary information in the brackets has been deleted in this report, the deleted information is provided in the proprietary version of this report (LTR-PAFM-16-59-P Revision 1).

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SVP_SV0_004439 ND-22-0817 Enclosure I - VEGP 3&4-ISil -ALT-17 Attachment 2 Westinghouse Non-Proprietary Class 3 Page 4 of 22 Page 14 of 38 LTR-PAFM-16-59-NP Revision 1 1.0 Introduction The objective of this letter report is to provide responses to the NRC Request for Additional Information (RAI) [l]

regarding the AP1000 Steam Generator (SG) to Reactor Coolant Pump (RCP) suction nozzle dissimilar metal (DM) weld inspection coverage. The NRC RAI requests additional information or analyses to justify why an ultrasonic examination of the inner 1/3 of the weld thickness and a surface examination of the inner and outer weld surfaces is an acceptable alternative to examining the full weld volume.

The responses to the NRC RAI will be based on two separate assessments that have been performed for the region of interest. The first assessment is based on an ASME Section XI flaw tolerance analysis, and the second assessment is based on the ASME Section III design evaluation.

1.1 ASME Section XI Flaw Tolerance Evaluations for Flaws on the Outside Surface and Embedded Within the Weld Examination Volume The first evaluation is based on an ASME Section XI flaw tolerance analysis for the DM weld, with the consideration of a surface postulated flaw size in the outer 2/3 of the wall thickness . This flaw evaluation considers a crack growth calculation for 60 years (design life) and the maximum end-of-evaluation flaw size determinations based on limit load analysis, typical of an ASME Section XI flaw evaluation using the rules of Appendix C. The intent is to show that a large postulated outside surface flaw will not grow to the maximum allowable end-of-evaluation period flaw size (i.e., allowable flaw size) after 60 years of growth. The maximum allowable end-of-evaluation period flaw size is calculated based on the ASME IWB-3640 guidelines [2]. The postulated outside surface flaws used in this crack growth analysis are an axial flaw with Aspect Ratio (AR), flaw length/flaw depth, AR= 2, and a circumferential flaw with AR= 6. The aspect ratio of 2 is reasonable because the axial flaw growth is limited to the width of the DM weld configuration, and an aspect ratio of 6 for postulated circumferential flaw is typical for fracture mechanics analyses. The primary crack growth mechanism for flaws within the nozzle welds is Fatigue Crack Growth (FCG). The fatigue crack growth rates as well as the stress intensity factor equations required to complete a FCG analysis are further discussed in Section 2 of this letter report. Crack growth due to Primary Water Stress Corrosion Crack (PWSCC) growth does not need to be investigated since the base metals around the DM weld, the stainless steel buttering, and the Alloy 152/52 DM weld material have a low susceptibility to stress corrosion cracking. Furthermore, the evaluation considered in this letter is for postulated outside surface flaws (which conservatively bound postulated embedded fabrication flaws) which are not exposed to the primary coolant, and thus the susceptibility to PWSCC is not of concern. Any potential indications in the inner 1/3 of the dissimilar metal weld wall thickness will be detected by volumetric inspection during the in-service inspections .

While the fracture mechanics evaluation assumed postulated outside surface flaws , the eight (8) steam generator to RCP casing DM welds for Vogtle Units 3 and 4, and six (6) of the eight (8) steam generator to RCP casing DM welds for V. C. Summer Units 2 and 3 (Note: V.C. Summer Unit 3 Steam Generator ' A' is still in the fabrication facility) have been examined in accordance with the requirements of ASME Section III, NB-5000. These examinations included a liquid penetrant (PT) examination of the outside surface. No relevant indications were observed. Relevant indications are those having major dimensions greater than 1/16-inch (0.0625-inch). For the remaining two (2) DM welds for V.C. Summer Unit 3, there can be no cracks or linear indications greater than Page 4 of22

SVP _SV0_004439 ND-22-0817 Enclosure I - VEGP 3&4-1S1 I-ALT- 17 Attachment 2 Westinghouse Non-Proprietary Class 3 Page 5 of 22 Page 15 of 38 LlR-PAFM-16-59-NP Revision 1 1/16-inch (0 .0625-inch) long or rounded indications greater than 3/ 16-inch (0 .188-inch) long on the outside surface.

This same evaluation of postulated outside surface flaws conservatively covers evaluations for embedded fabrication flaws which may be present at the beginning of service. The stress intensity factors in the fracture mechanics analysis for surface flaws are more limiting than embedded fabrication flaws.

The eight (8) steam generator to RCP casing DM welds for Vogtle Units 3 and 4, and six (6) of the eight (8) steam generator to RCP casing DM welds for V.C. Summer Units 2 and 3 have been examined volumetrically in accordance with the requirements of ASME Section III, NB-5000 and Westinghouse design requirements. These volumetric examinations included radiographic (RT) and ultrasonic (UT) testing of the buttering materials on both the steam generator and RCP materials as in-process examinations, and RT and UT of the weld, buttering materials and 0.25-inch of adjacent base material on both sides of the weld for the full thickness after completion of the welds . The post-weld RT results were evaluated against the acceptance standards in ASME Code Section III, NB-5320 . The post-weld UT of the full volume of the weld used multiple angle, four directional angle beam techniques, from both the ID (inside diameter) and OD (outside diameter) surfaces. The post-weld UT results were evaluated against the acceptance standards of ASME Code Section III, NB-5331 and Section XI, IWB-3514 (for preservice examination).

The UT from the OD surface was performed with a 0° probe and 45°, 60°, and 70° 1.0 MHz transmit-receive, longitudinal wave probes focused at various depths below the OD surface; the angle beam examinations were focused on the outer half of the examination volume. The UT from the ID surface was performed with 37°, 45°,

and 70°, 1.0 - 2.0 MHz transmit-receive longitudinal wave probes focused at various depths below the ID surface; these examinations were focused on the inner half of the examination volume .

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SVP_SV0_004439 ND-22-0817 Enclosure l - VEGP 3&4-ISil -ALT-17 Attachment 2 Westinghouse Non-Proprietary Class 3 Page 6 of 22 Page 16 of 38 LTR-PAFM-16-59-NP Revision 1 The remaining two (2) steam generator to RCP casing DM welds in the V.C. Summer Unit 3 SG ' A' that have not been examined may contain volumetric RT flaw indications of less than 0.75-inch long . If there are any UT flaw indications these flaw indications must satisfy the allowable flaw standards in ASME Code Section XI 1998 Edition with the 2000 Addenda, Table IWB-3514-2 for preservice examination with a nominal wall thickness of 3.0-inches. For a subsurface flaw indication, the allowable a/t ranges from 7.6% to 8.9% depending on the aspect ratio (all) of the flaw indication. This converts to a through-wall extent (2a/t) of 15 .2% to 17.8%, or for a [

tc.e Therefore, the allowable standards in Table IWB-3514-2 are sufficiently large to show acceptance for any flaws detected during fabrication and pre-service at Vogtle and V. C. Summer. It should be noted that the ASME Section XI 2007 Edition, 2008 Addenda have even larger allowable standards than those in the 1998 Edition with 2000 Addenda.

1.2 ASME Section III Design Evaluations In addition to the ASME Section XI flaw tolerance analysis for postulated axial and circumferential flaw on the outside surface (and postulated embedded fabrication flaws) of the SG to RCP suction nozzle DM weld, a Section III design evaluation [4] assessment was already completed for the AP 1000 Steam Generator and the adjacent DM weld. The primary goal of the Section III evaluation is to demonstrate acceptable margins to avoid cracks initiating as a result of fatigue cycling. The ASME Section III discussion and results are provided here also to demonstrate that the region of interest (i .e., SG outlet nozzle and DM weld) meet the structural design requirements of the ASME Code. The select results that are provided in Section 3 of this letter aim to demonstrate that the primary and secondary stress analysis, fatigue usage, and non-ductile failure (fracture mechanics) assessment per the ASME Section III code are all satisfied. Therefore, meeting the requirements of ASME Section III further demonstrates the justification that the DM weld location is structurally qualified for the design life of the plant.

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SVP_SV0_004439 ND-22-0817 Enclosure 1 - VEGP 3&4-ISII-ALT-17 Attachment 2 Westinghouse Non-Proprietary Class 3 Page 7 of 22 Page 17 of 38 LTR-PAFM-16-59-NP Revision 1 2.0 ASME Section XI Flaw Tolerance Analysis for Postulated Outside Surface (and Embedded) Flaws This section provides a brief discussion for the fracture mechanics analysis of outside surface postulated flaws per the ASME Section XI guidelines. The evaluation first calculates the maximum allowable end-of-evaluation period flaw sizes for the two different flaw orientations (axial and circumferential flaws) based on ASME Section XI at the Steam Generator to RCP DM weld location. Next, fatigue crack growth calculations at the dissimilar metal weld are performed for 60 years for large postulated outside surface flaws . The initial postulated outside surface flaw sizes are sufficiently larger than any existing fabrication-related outside surface flaws detected during the fabrication examinations of the welds.

It is also noted that since the stress intensity factor is lower for embedded flaws than that of surface flaws for a given through-wall stress distribution, the initial postulated outside surface flaw sizes are bounding for existing fabrication-related embedded flaws .

2.1 Maximum Allowable End-of-Evaluation Period Flaw Sizes The calculation of the maximum allowable end-of-evaluation period flaw sizes for austenitic steel and nickel base alloys is based on limit load analysis . The procedures and acceptance criteria for the limit load analysis in austenitic components and weld metals are contained in paragraph IWB-3640 of ASME Section XI [2]. These criteria were used to determine the maximum allowable end-of-evaluation period flaw size for axial (AR= 2) and circumferential (AR= 6) flaw configurations. The aspect ratio of 2 is reasonable because the axial flaw growth is limited to the width of the DM weld configuration, and an aspect ratio of 6 for postulated circumferential flaw is typical for fracture mechanics analyses . The procedure to evaluate the allowable flaw sizes is based on IWB-3640 and subsequently Appendix C of Section XI of the code.

The maximum end-of-evaluation period flaw sizes determined for both axial and circumferential flaws have incorporated the relevant material properties, nozzle loadings and geometry. Loadings under normal, upset, emergency, and faulted conditions were considered in conjunction with the applicable safety factors for the corresponding service conditions required in the ASME Code Section XI. For circumferential flaws, axial stress due to the [

tc.e were considered in the evaluation. As for the axial flaws ,

hoop stress resulting from pressure loading was used, since none of the other loadings have an impact on such flaws .

Per ASME Section XI, the thermal expansion loads do not need to be considered in the maximum end-of-evaluation period flaw size determination since the nozzle welds are GTAW (Gas Tungsten Arc Weld) and are non-flux welds .

The AP 1000 SG to RCP suction nozzle DM weld dimensions and operating parameters are shown in Table 1. A design temperature of [ ]ac.e was conservatively used in determining the end-of-evaluation period flaw size and for the fatigue crack gro-wth analysis. The nozzle loads at the SG to RCP suction nozzle weld are based on conservatively bounding both the SG and RCP design specification allowable loads (Table 2) . The loads given in Table 2 are in the local coordinate system, where the x-axis is axial along the component centerline, y-axis and z-Page 7 of22

SVP_SV0_004439 ND-22-0817 Enclosure 1 - VEGP 3&4-ISIJ-ALT-17 Attachment 2 Westinghouse Non-Proprietary Class 3 Page 8 of 22 Page 18 of 38 LTR-PAFM-16-59-NP Revision 1 axis by right-hand-rule. Furthermore, all loads are conservatively applied as absolute values. The design mechanical loads cover normal pump vibration loadings.

Table 1: AP1000 SG to RCP Suction Nozzle Weld Geometry and Operating Parameters a,c,e Table 2: AP1000 SG to RCP Suction Nozzle Weld Allowable Loads a,c,e Page 8 of22

SVP_SV0_004439 ND-22-0817 Enclosure I - VEGP 3&4-ISII -ALT-17 Attachment 2 Westinghouse Non-Proprietary Class 3 Page 9 of 22 Page 19 of 38 LTR-PAFM-16-59-NP Revision 1 The maximum end-of-evaluation period allowable flaw sizes are determined based on the weaker of the base metal and weld metal material properties flow strength value (average of the yield and ultimate strengths) at the SG to RCP suction nozzle weld for a maximum temperature of [ yc.e_ The ASME code limiting material properties at the DM weld location are based on the [

The maximum allowable end-of-evaluation period flaw sizes for the SG to RCP suction nozzle DM weld are shown in Table 3. It should be noted that the maximum end-of-evaluation period allowable flaw sizes are limited to only 75% of the wall thickness in accordance with the requirements of ASME Section XI paragraph IWB-3640

[2]. Next, the fatigue crack growth analyses are performed to determine the largest postulated allowable initial flaw size for 60 years of plant operation such that the final crack growth flaw sizes will not reach the maximum-end-of-evaluation period flaw sizes shown in Table 3.

Table 3: Maximum Allowable End-of-Evaluation Period Flaw Size Aspect Ratio (flaw Maximum End-of-Evaluation Flaw Configuration length/flaw depth) Period Flaw Size (alt)

Axial Flaw 2 0.75 Circumferential Flaw 6 0.47 The wall thickness, denoted as 't' , and the flaw depth and flaw length, denoted as ' a ' and 'e' respectively, are shown in Figure 1 for an axial flaw on the outside diameter of the SG to RCP suction nozzle DM weld. A circumferential flaw on the outside diameter has the same denotation for thickness and flaw configuration variables.

7 Inside Diame1er Flaw Thickness J

(t)

Outside Flaw Depth Diameter (a)

....._________.~

L- Flaw Length __J 1* c:i I Figure 1: Illustration of SG to RCP Suction Nozzle DM Weld Outside Diameter Axial Flaw Page 9 of22

SVP_SV0_ 004439 ND-22-0817 Enclosure 1 - VEGP 3&4-ISII-ALT-17 Attachment 2 Westinghouse Non-Proprietary Class 3 Page 10 of 22 Page 20 of 38 L TR-P AFM-16-59-NP Revision 1 2.2 Crack Growth Evaluations As discussed in Section 1 of this letter report, the primary crack grmvth mechanism for flaws within the nozzle welds is fatigue crack growth. Crack growth due to Primary Water Stress Corrosion Crack growth is not applicable since the Alloy 152/52 DM weld material has very low susceptibility to stress corrosion cracking, especially for outside surface flaws and embedded flaws which are not exposed to the primary coolant. The fatigue crack growth rate as well as the stress intensity factor equations required to complete a FCG analysis are discussed in this section. The inputs that are used in the crack growth evaluations are the applicable design transients and cycles that contribute significantly to the FCG (see Section 2.2.1), along with the consideration of welding residual stresses to the crack grmvth analysis. The through-wall transient stresses, along with the residual stresses are used to calculate the stress intensity factor, which will then be used with the cycles to determine the 60 year FCG results for the axial and circumferential flaw.

2.2.1 Transient Stresses The transient hoop and axial stress data used in the FCG analysis is representative of the stresses at the SG to RCP suction nozzle DM weld. The transients and cycles are listed in Table 4, and these particular transients are considered limiting for use in the FCG analysis . The remaining AP 1000 design transients do not have high fatigue usage factor contributions to the AS.ME Section III RCP nozzle fatigue evaluation. It is assumed that the fatigue crack growth is also negligible for the other remaining design transients, since the fatigue usage factor contribution for those transients is negligible . It should be noted that the mechanical loadings are included in the transient stresses. A high-cycle pump vibration transient has also been considered, and determined to have stress intensity factor that are below the threshold needed for fatigue crack growth as described later in this report (Section 2.2.5).

Table 4: Limiting Transients Used for the FCG Analysis a,c,e Page 10 of 22

SVP_SV0_004439 ND-22-0817 Enclosure 1 - VEGP 3&4-ISII -ALT- 17 Attachment 2 Westinghouse Non-Proprietary Class 3 Page 11 of 22 Page 21 of 38 LTR-PAFM-1 6-59-NP Revision 1 2.2.2 Residual Stresses For the FCG analysis, the welding residual stresses at the SG to RCP suction nozzle DM weld are also considered along with the transient stresses. The inclusion of residual stresses will not change the range of stress intensity factor for the fatigue crack growth calculations; however, it will affect the Load Ratio (R) in the FCG equation (see Section 2.2.5). [

2.2.3 Fatigue Crack Growth Analysis In order to determine the growth of a postulated flaw after 60 years, a fatigue crack growth analysis is completed.

Fatigue crack growth is the only credible mechanism for crack growth in the material between the SG and RCP since both the weld and the base metals have very low susceptibility to PWSCC, especially since the outside postulated surface flaw is not exposed to the reactor coolant loop environment. The fatigue crack gro\vth analysis procedure involves postulating an initial flaw at the weld region and predicting the growth of that flaw due to an imposed series of loading transients. The input required for a fatigue crack growth analysis is essentially the information necessary to calculate the range of crack tip stress intensity factor, which depends on the crack size and shape, geometry of the structural component where the crack is postulated, and the applied cyclic stresses.

Provided below is the methodology used to calculate the stress intensity factor for the axial and circumferential surface flaws .

2.2.4 Generation of Crack Tip Stress Intensity Factors The FCG analysis in this letter involves calculating growth for a flaw on the outside surface of the SG to RCP suction nozzle DM \Veld, for an axial (AR = 2) and circumferential (AR = 6) flaw. The aspect ratio of 2 is reasonable because the axial flaw growth is limited to the width of the DM weld configuration, and an aspect ratio of 6 for postulated circumferential flaw is typical for fracture mechanics analyses. The postulated flaws are subjected to cyclic loads due to the transients and residual stresses described previously. The inputs required for the fatigue crack growth analysis is the range in stress intensity factor, t-K, and the R ratio, KmiiJKmax-The stress intensity factors expression for surface flaws utilizes a representation of the actual stress profile rather than a linearization between data points. The stress distribution profiles are represented by a cubic polynomial:

where:

A0 , A 1, A 2 , and A3 are the stress profile curve fitting coefficients, x is the distance from the wall surface where the crack initiates, and cr is the stress perpendicular to the plane of the crack.

Page 11 of 22

SVP_SV0_004439 ND-22-0817 Enclosure 1 - VEGP 3&4-ISil-ALT-17 Attachment 2 Westinghouse Non-Proprietary Class 3 Page 12 of 22 Page 22 of 38 LTR-PAFM-16-59-NP Revision 1 The stress intensity factor expression for semi-elliptical flaw shapes was used. The methodology for calculating the crack tip stress intensity factors is documented in an ASME publication [5] for axial flaws. The stress intensity factor from [5] can also be used conservatively for circumferentially oriented flaws . When evaluating axial and circumferential flaws , semi-elliptical surface flaws with aspect ratios (flmv length/flaw depth) of 2 for axial flaws and 6 for circumferential flaws are considered. Stress intensity factors can be expressed in the general form as follows:

3

_ na o.s " ~

K1 - ( - ) .,..uj (ale, alt, t!R, <!>) Ai aj Q j=O where:

a: crack depth c: half of the crack length along the surface wall thickness inside radius of the component coefficients Ao, A 1, A 2, and A 3 for the stress profile cubic fit angular position of a point of the crack front

(~= 0° at the deepest point; 90° at the surface point)

Go, G1, G2, G3 are boundary correction factors provided Ill [5] for axial and used conservatively for circumferential flaws Q: shape factor of an elliptical crack. Q is approximated by:

Q = 1 + l.464(alc)1 65 for ale :-:::1, or Q = 1 + l .464(c/a}1- 65 for ale> l 2.2.5 Fatigue Crack Growth Rate Once R (load ratio = Knun1Kmax) and £1.K are calculated, the crack growth due to any given stress cycle can be calculated for each transient. This increment of crack grmvth is then added to the original crack size, and the analysis proceeds to the next transient.

Fatigue crack grmvth for each transient for a given time interval and number of cycles (N) can be computed using the following equation:

New Crack Depth= Initial Crack Depth+ Incremental Crack Depth with the incremental crack depth, .6a, given by:

The procedure is continued in this manner until all the transients known to occur in the period of evaluation have been analyzed. The design transient load cycles used in the analysis for the AP1000 SG to RCP suction nozzle Page 12 of 22

svP_svo_oo4439 ND-22-0817 Enclosure I - YEGP 3&4-ISII-ALT-17 Attachment 2 Westinghouse Non-Proprietary Class 3 Page 13 of 22 Page 23 of 38 LTR-PAFM-16-59-NP Revision 1 weld are listed in Table 4. The above equation is the most fundamental form of fatigue crack growth law, where C and n are material constants.

The general fatigue crack growth rate for materials in air environments is given by the equation of the type:

da 11 dN = Fweld C(T) S(R) (L'.lK) where:

C(T) Scaling Factor for Temperature Effects S(R) Scaling Factor for Load Ratio Effects Factor for Weld Material Stress Intensity Factor Range = Kmax - Kmiu R Load Ratio Kmin / Kmax Maximum Stress Intensity Factor Minimum Stress Intensity Factor da/dN Crack Growth Rate in Environment n Crack Growth Law Exponent The fatigue crack growth is performed for the Alloy 152/52 \veld material between the SG and RCP suction nozzle . Note that the buttering on the steam generator outlet nozzle is Alloy 152/52, and the weld is Alloy 52. The FCG reference curves for Alloy 152/52 have not been developed in Section XI of the ASME Code; therefore, information available in NUREG/CR-6907 [6] is used. According to [6] , in an air environment the Alloy 52 and Alloy 182 weld is approximately 2 times the Alloy 600 FCG rate in air. Due to limited number of test data for Alloy 152 in air environment, Reference [6] concludes that a factor of 2 on the Alloy 600 in air can be used to approximate the Ni-alloy welds, such as Alloy 152/52, FCG rate in air. It should be noted that the buttering on the RCP pump suction nozzle is stainless steel; however, the crack growth results for the Ni-alloy in air are more limiting than the stainless steel material in air (FCG curves for stainless steel based on ASME Section XI Appendix C).

Thus, the crack growth evaluation used herein are based on the FCG rate expression for Alloy 600 in air in SI units with a factor of 2 to represent the Alloy 152/52 weld in air environment [6]:

da 11 dN = Fweld C(T) S(R) ( L'.lK) qn = 4.835 X 1o*l-l + (1.622 X 10"16)T- (1.490 X 10"18 )T2 + (4.355 X 10"21 )T3 S(R)= (1 - 0.82Rr2 2 Page 13 of22

SVP_SV0_004439 ND-22-0817 Enclosure I - VEGP 3&4-ISil -ALT- 17 Attachment 2 Westinghouse Non-Proprietary Class 3 Page 14 of 22 Page 24 of 38 LTR-PAFM-16-59-NP Revision 1 Fweld =2 where :

T = Temperature (0 C)

Af( = Stress Intensity Factor Range, Kmax - Kmin, MPa m_

Kmax = Maximum Stress Intensity Factor, MPa m_

Kmin = Minimum Stress Intensity Factor, MPa m_

n = Crack Growth Law Exponent(= 4.1)

R = Load Ratio, Kmin / Kmax da

= Crack Growth Rate in Environment, m/Cycle dN

]a.c.e As such, the stress profile and stress range through the DM weld thickness due to RCP vibrations vvill be small. The stresses that are produced by the vibration are below the endurance limit of the Alloy 152/52 DM weld.

Furthermore, the range in stress intensity factors for pump vibrations are less than the ~Kttrreshotd

Therefore, the contribution of RCP vibrations to the FCG analysis would be negligible.

Page 14 of 22

SVP_SV0_004439 ND-22-0817 Enclosure 1 - VEGP 3&4-ISII-ALT-17 Attachment 2 Westinghouse Non-Proprietary Class 3 Page 15 of 22 Page 25 of 38 LTR-PAFM-16-59-NP Revision 1 2.2.6 Fatigue Crack Growth Charts The fatigue crack growth charts (Figures 2 and 3) are constructed by plotting the fatigue crack growth results over a period of 60 years. The flaw depth to through-wall thickness ratio (alt) is plotted as the ordinate, and time is plotted as the abscissa. The charts are generated for the SG to RCP suction nozzle DM weld for an outside surface axial flaw (AR= 2) and circumferential flaw (AR= 6) as shown in Figures 2 and 3 respectively. The fatigue crack growth results are compared to the maximum allowable end-of-evaluation flaw size. The maximum allowable flaw size is tabulated in Table 3 for the axial and circumferential flaws , and also plotted in Figures 2 and 3. The initial flaw size is a sufficiently large postulated flaw which would not reach the maximum allowable flaw size in 60 years.

As shown in Figure 2 for an axial flaw, even a 60 percent through the wall thickness flaw would not reach the maximum allowable flaw size in 60 years. Figure 3 for circumferential flaws shows that a postulated flaw as large as 30 percent through the wall thickness flaw would not reach the maximum allowable flaw size in 60 years. Any initial axial and circumferential flaw sizes less than the 60 and 30 percent of the wall thickness, respectively, are encompassed by these curves and wi ll not grow to the maximum allowable flaw size in 60 years. The large axial and circumferential surface flaw sizes described above do not exist in the eight (8) DM welds of Vogtle Units 3 and 4, and six (6) of the eight (8) DM welds ofV.C. Summer Units 2 and 3 as evidenced by the ASME Section III fabrication PT examination results. The remaining two (2) DM welds in V.C. Summer Unit 3 will not contain surface flaws as described above given the allowable standards of ASME Code Section III, NB-5000 . Surface flaws will be detected by the regular ISi surface examinations of the outside surface of the SG to RCP suction nozzle DM weld.

The stress intensity factor correlations for an embedded flaw are lower than that for surface flaws. Therefore, for a given through-wall stress distribution, it can be concluded that the fatigue crack growih for outside surface flaws is more limiting than the embedded flaws due to higher stress intensity factor. [

rc.e Thus, all the UT detected indications are below the ASME Section XI IWB-3500 Allowable Standards. Furthermore, the initial surface flaw sizes used in the crack growth analysis in Figures 2 and 3 bound not only all the UT detected indications in the existing SG to RCP welds that have been inspected, but also bound the ASME Section XI allowable flaw sizes. Therefore, for the remaining two V. C. Summer Unit 3 SG A DM welds that have not been inspected, it can be conservatively assumed to have initial embedded flaw sizes (alt) no larger than the ASME Section XI allowable flaw sizes in the (2a/t) range of 15.2% (0.845-inch) to 17.8% (0 .99-inch) of the wall thickness per ASME Section XI Table IWB-3514-2 (1998 edition 2000 addenda). Thus, the initial surface flaw sizes used in Figures 2 and 3 will also bound the remaining welds that are not yet inspected.

In conclusion, all detected indications and any other potential fabrication indications are within the ASME Section XI allowable standards (1998 edition 2000 addenda), and moreover bounded by the initial flaw sizes considered in the crack growth analysis performed in this report for the design life of the plant (60 years).

Page 15 of 22

SVP_SV0_004439 ND-22-0817 Enclosure I - VEGP 3&4-ISII -ALT- 17 Attachment 2 Westinghouse Non-Proprietary Class 3 Page 16 of 22 Page 26 of 38 LTR-PAFM-16-59-NP Revision 1 0.8

~----- *- - I ASME Maximum Allowable End-Of-Evaluation Period Flaw Size ~-..,--------

0.7 I 0.6 I I I I I I

0.5

~

<l,j

.§ 0.4 t.l

a E--<

=a

~

-=i:::..

~

<l,j 0.3

~

f;;; 0.2 I

0.1 I

0 0 LO 20 30 40 50 60 Service Life (year)

Figure 2: Crack Growth Chart for the AP1000 Steam Generator to Reactor Coolant Pump Suction Nozzle Dissimilar Metal Weld, Outside Surface Axial Flaw with Aspect Ratio = 2 Page 16 of 22

ND-22-0817 Enclosure I - VEGP 3&4-ISII-ALT- 17 SVP_SV0_004439 Westinghouse Non-Proprietary Class 3 Page 17 of 22 Page 27 of 38 LTR-PAFM-16-59-NP Revision 1 0.8 , - - - - - - - - -- - - - - - - - - , - - - - - - - - - - -- -- - --------,

0.7 +-- - - - - - - - - -*- - - - - - - - - - - + - - - - - - - - - - -- ---

0.6 --+

0.5

~ - - - - - - - ASME Maximum Allowable End-Of-Evaluation Period Flaw Size - - - - - - - - - -

~ 0.4 + - - - - - - - - - - -*- - - - - - - - - - - -

i

.c C.

~"'

0.3 L~::::::t::=:===-==-==-==-=

£ 0.2 + - - - - - - - - - -

0. 1 0

0 10 20 30 50 60 Sen-ice Life (.vear)

Figure 3: Crack Growth Chart for the AP1000 Steam Generator to Reactor Coolant Pump Suction Nozzle Dissimilar Metal Weld, Outside Surface Circumferential Flaw with Aspect Ratio = 6 Page 17 of 22

SVP_SV0_004439 ND-22-0817 Enclosure 1 - VEGP 3&4-ISII-ALT-17 Attachment 2 Westinghouse Non-Proprietary Class 3 Page 18 of 22 Page 28 of 38 LTR-PAFM-16-59-NP Revision 1 3.0 Section III Design Evaluation for Steam Generator to Pump DM Weld The goal of the discussion herein on ASME Section III design evaluation [4] is to supplement the primary assessment provided in Section 2 based on ASME Section XI flaw tolerance analysis. The aim here is to provide a brief summary of the primary and secondary stress analyses, including the fatigue usage and ASME Section III Appendix G fracture mechanics (low alloy ferritic steel region) results.

The ASME Section III evaluations for the SG primary outlet nozzle and the SG to RCP suction nozzle DM weld are based on the pressure loads, thermal loads, and external mechanical loads obtained using finite element analysis and also based on strength of materials equations . It should be noted that the loads due to pump fluctuations and vibrations were included in the evaluation for all conditions to determine the fatigue usage factors. Furthermore, the high cycle loading due to pump vibrations was also evaluated separately for an infinite number of cycles to determine the maximum alternating stress. All alternating stress intensities for this high cycle pump loading are below one-half the endurance limits.

Provided in Table 5 below are select results of the ASME Section III allowable stress limits and fatigue usage for the DM weld. Note that all ASME Section III design criteria are satisfied for this region . Furthermore, the low fatigue usage shown in Table 5 demonstrates low susceptibility for any fatigue crack initiations at either inside or outside surfaces.

Table 5: ASME Section III Select Results for DM Weld Location a,c,e A non-ductile fracture mechanics evaluation was also performed per ASME Section III Appendix G for the SG nozzle ferritic material adjacent to the DM we ld. The non-ductile brittle fracture evaluation per ASME Section III Appendix G can be used as a conservative fracture mechanics assessment of the more ductile Alloy 152/52 weld.

The Appendix G results for the ferritic location in the SG next to the DM weld are shown in Table 6 below.

Page 18 of 22

SVP_SV0_004439 ND-22-0817 Enclosure l -VEGP3&4-ISI1-ALT-17 Attachment 2 Westinghouse Non-Proprietary Class 3 Page 19 of 22 Page 29 of 38 LTR-PAFM-16-59-NP Revision 1 Results from Table 6 can be used to demonstrate the structural stability of the region at and around the DM weld based on a large postulated flaw of 25 percent of the wall thickness with a length-to-depth (aspect ratio) of 6, as required by ASME Section III Appendix G design analysis. Normal, upset, emergency, test, and faulted condition transients were all evaluated. The limiting transients within these service conditions were chosen to be those transients that result in low metal temperatures and high stresses, which would give the worst brittle fracture assessment. The most limiting hoop and axial stresses from either the inside or outside surface were used in the Appendix G evaluation; as a result, the evaluation cover postulated flaws on the inside or outside, and axial or circumferential flaw configurations . The lower bound fracture toughness values were used based on the limiting temperature and material reference nil-ductility temperature (RTNDT) based on the design specification. Table 6 provides the ASME Section III results for the postulated l/4T flaw for the ferritic steel location adjacent to the DM weld. Based on the results in Table 6, it is demonstrated that the ferritic steel and the adjacent ductile DM weld would be flaw tolerant for large postulated flaws based on the ASME Section III Appendix G non-ductile evaluations.

Table 6: ASME Section III Appendix G Results' for SG Nozzle Ferritic Steel Location Adjacent to the DM weld a,c,e Page 19 of 22

SVP_SV0_004439 ND-22-0817 Enclosure I - VEGP 3&4-ISII-ALT-17 Attachment 2 Westinghouse Non-Proprietary Class 3 Page 20 of 22 Page 30 of 38 LTR-PAFM-16-59-NP Revision 1 4.0 Conclusions The objective of this letter report is to provide responses to the NRC RAI (Reference 1) to support justification of a volumetric inspection of the inner 1/3 of the weld, with a surface examination of the inner and outer weld surfaces by ET and PT, respectively, and to demonstrate that this is an acceptable alternative to examining the full weld volume during the ISi. The responses to the NRC RAI were based on two separate assessments that have been performed for this particular region of interest. The first assessment was based on ASME Section XI flaw*

tolerance analysis, and the second assessment is based on the ASME Section III design evaluation.

The ASME Section XI flaw tolerance evaluation is provided in Section 2 of this report. Postulated outside surface axial and circumferential flaws with aspect ratios of 2 and 6, respectively, were evaluated at the SG to RCP suction nozzle DM weld locations . APl000 specific geometry, loadings, and material properties were considered in the maximum end-of-evaluation period flaws and the fatigue crack growth analysis. [

] a.c.e Thus, all detected indications are well within the ASME Section XI allowable standards and the fracture mechanics calculations performed in this report.

Section 3 of this report provides the ASME Section III design evaluation that *was performed for the DM weld and the surrounding low alloy steel region. Based on the design criteria, all requirements of the ASME Section III code were met for this region based on the primary and secondary stress analyses, and fatigue usage calculations.

The fatigue usage at the DM weld region is very low at both the inside and outside surface, and this region has acceptably low susceptibility to crack initiations for the design life of the plant. The non-ductile ASME Section III Appendix G fracture mechanics evaluation was also performed and shown to be acceptable for the low alloy ferritic steel of the steam generator. The ferritic steel material is considered in the Appendix G evaluation since it is susceptible to brittle fracture , whereas the DM weld is more ductile than the SG base metal, and therefore is not required to be considered in the Appendix G evaluation. Thus, per the design ASME Section III fracture mechanics, it is also demonstrated that the DM we ld region is flaw tolerant for large flaws of size 25% of the wall thickness with an aspect ratio of 6.

Page 20 of22

SVP _SV0_004439 ND-22-0817 Enclosure 1 - VEGP 3&4-ISII-ALT-17 Attachment 2 Westinghouse Non-Proprietary Class 3 Page 21 of 22 Page 31 of 38 LTR-PAFM-16-59-NP Revision 1 In conclusion, based on the ASME Section XI and III discussions provided in this report, it is demonstrated that with the volumetric and surface examinations performed during fabrication, and the outer surface examinations and inner 1/3 of the wall thickness surface and volumetric examinations to be performed every ISi, a volumetric inspection of the entire DM weld that includes the outer 2/3 of the wall thickness is not necessary. This conclusion is based on a fracture mechanics evaluation per ASME Section XI and ASME Section III Appendix G assessments. It was demonstrated that the outer 2/3 of the wall thickness is flaw tolerant for the design life of the plant, and that the initiation of any active degradation of the weld would not be expected to occur over the licensed lifetime of the plant.

Page 21 of22

SVP_SV0_004439 ND-22-0817 Enclosure l - VEGP 3&4-ISII -ALT-17 Attachment 2 Westinghouse Non-Proprietary Class 3 Page 22 of 22 Page 32 of 38 L1R-PAFM-16-59-NP Revision 1 5.0 References

1) NRC email from Steven Downey to Chandu Patel, "Requests for Additional Information related to Vogtle Alternative Request VEGP3&4-PSI-ALT-05 ," dated: August 5, 2016, (NRC ADAMS: ML16218A439) .

2) ASME Code Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components," 2007 Edition including 2008 Addenda.

3) Request for Alternative Requirement for Preservice Inspection at Vogtle Units 3 & 4 and V.C. Summer Units 2 and 3.

a. VEGP 3&4-PSI-ALT-05, "Southern Nuclear Operating Company, Vogtle Electric Generating Plant Units 3 and 4 Request for Alternative: Preservice Inspection Requirements for Steam Generator Nozzle to Reactor Coolant Pump Casing Welds," June 24, 2016, (NRC ADAMS : ML16176A312).

b. NND-16-0246, "Virgil C. Summer Nuclear Station (VCSNS) Units 2 and 3, Docket Numbers52-027 and 52-028, Request for Alternative: Preservice Inspection Requirements for Steam Generator Nozzle to Reactor Coolant Pump Casing Welds," July 7, 2016, (NRC ADAMS: ML16189A312).

4) ASME Code Section III, "Rules for Construction of Nuclear Power Plant Components," 1998 Edition including 2000 Addenda.

5) Raju, I.S. and Newman, J.C. , "Stress Intensity Factor Influence Coefficients for Internal and External Surface Cracks in Cylindrical Vessels," ASME Publication PVP, Volume 58, 1982, pp. 37-48.

6) NUREG/CR-6907, ANL-04/3 "Crack Growth Rates of Nickel Alloy Welds in a PWR Environment," U.S.

Nuclear Regulatory Commission (Argonne National Laboratory), May 2006.

7) NUREG/CR-6383, ANL-95/37, "Corrosion Fatigue of Alloys 600 and 690 in Simulated LWR Environments," April 1996.

8) Nomura, Y., Yamamoto, K. , Hojo, K. , ASME PVP2014-28098, "Fatigue Crack Growth Rates for Nickel Base Alloys in Air," Proceedings of ASME 2014 Pressure Vessels & Piping Conference, Anaheim, California, USA, July 20-24, 2014 .

9) U.S. Nuclear Regulatory Commission Letter Dated May 19, 2000, License Renewal Issue No . 98-0030,

Thermal Aging Embrittlement of Cast Austenitic Stainless Steel Components."

10) Doosan Radiographic Test Reports for Vogtle Unit 3 SG A (Rl50209-017-001 and Rl50209-020-001) and SG B (Rl41201-043-001 and Rl41201-044-001) and for Vogtle Unit 4 SG A (Rl50430-005-001 and Rl50430-006-001) and SG B (Rl50709-015-001 and Rl50709-016-001) .

11) Doosan Ultrasonic Test Reports for Vogtle Unit 3 SG A (Ul50212-042-001) and SG B (Ul41201-029-001) and Vogtle Unit 4 SG A (Ul50417-005-001) and SG B (Ul50709-036-001).

12) Doosan Radiographic Test Reports for V.C. Summer Unit 2 SG A (Rl40805 -021-001 and Rl40805-022-001) and SG B (Rl40902-014-001 and Rl40901-020-001) and for V.C. Summer Unit 3 SG B (Rl51 l 16-014-001 and Rl5ll16-015-001).

13) Doosan Ultrasonic Test Reports for V.C. Summer Unit 2 SG A (Ul40730-037-001) and SG B (U 140829-002-001) and for V.C. Summer Unit 3 SG B (Ul51119-012-001).

Page 22 of22

ND-22-0817 Enclosure 1 - VEGP 3&4-ISil -ALT-17 Page 33 of38 10CFR50.55a Alternative VEGP3&4-ISI1-AL T-17 Proposed Alternative in Accordance with 10CFR50.55a(z)(l)

--Alternative Provides Acceptable Level of Quality and Safety-Revision 0 Enclosure 2 Westinghouse LTR-PAFM-17-6, Rev. 0, NP-Enclosure 1 -Additional Information Regarding APlO00 Vogtle Units 3 & 4 and V. C. Summer Units 2 & 3 Steam Generator to Reactor Coolant Pump Suction Nozzle Weld Flaw Evaluation

SVP _SV0_004649 ND-22-0817 Enclosure 1- VEGP 3&4-ISII-ALT-17 Page 6 of 10 Page 34 of 38 Westinghouse Non-Proprietary Class 3 Page 1 of 5 LTR-PAFM-17-6 Rev. 0 NP-Attachment January 20, 2017 Attachment B NP-Attachment (Non-Proprietary)

Additional Information Regarding API0OO Vogtle Units 3 & 4 and V. C. Summer Units 2 & 3 Steam Generator to Reactor Coolant Pump Suction Nozzle Weld Flaw Evaluation This document contains Westinghouse Electric Company LLC proprietary information and data which has been identified by brackets . Coding (a.c.eJ associated with the brackets sets forth the basis on which the infonnation is considered proprietary. These codes are listed with their meanings in WCAP-7211 Revision 8 (September 2015), "Proprietary Information and Intellectual Property Management Policies and Procedures."

The proprietary information and data contained in this report we re obtained at considerable Westinghouse expense and its release could seriously affect our competiti ve position. This information is to be withheld from public disclosure in accordance with the Rules of Practice I0CFR2.390 and the information presented herein is to be safeguarded in accordance with I0CFR2 .903. Withholding ofthis infonnation does not adve rsely affect the public interest.

This information has been provided for your internal use onl y and should not be released to persons or organizations outside the Directorate of Regulation and the ACRS without the express written approval of Westinghouse Electric Company LLC. Should it become necessary to release this information to such persons as part of the re view procedure , please contact Westinghouse Electric Company LLC, which will make the necessary arrangements required to protect the Company's proprietary interests .

The proprietary information in the brackets has been deleted in this report. The deleted information is provided in the proprietary version of this report.

SVP_SV0_004649 ND-22-0817 Enclosure I - VEGP 3&4-ISII-ALT- 17 Page 7 of 10 Page 35 of 38 Westinghouse Non-Proprietary Class 3 Page 2 of 5 LTR-PAFM-17-6 Rev. 0 NP-Attachment January 20, 2017 Additional Information RegardingAPJOOO'.!l Vogtle Units 3 & 4 and V. C. Summer Units 2 & 3 Steam Generator to Reactor Coolant Pump Suction Nozzle Weld Flaw Evaluation

1. Additional infonnation regarding justification of the constant fl aw aspect ratios (AR) of 2 and 6 for axial and circumfe rential flaws in Refe rence 1 is provided below.

Axial Flaw For the postulated axial fl aw, the analysis in Reference I considers an AR = 2 (flaw length/flaw depth) . This flaw shape is based on the understanding that in the DM (dissimilar metal) weld, the axial fl aw will follo w the characteristic shape of the DM weld width and thickness . The DM weld inspection volume consists of the width of the dissimilar metal we ld and the Heat Affected Zone (HAZ) - see Figure 1. The width of the dissimilar metal weld is approximately [ t c.e based on the APl000 steam gene rator and pump drawings. The inspection volume includes the 1/4" examination zones adjacent to the we ld on either side to account for the HAZ . Therefore, the total width of the DM weld inspection region is approximately [ ]"-'*e_ The weld thickness is

[ ]a.c.e (Reference 1). Therefore the shape or aspect ratio of the weld is 0.6 [ ]a.c.c, thus an aspect ratio of 2 is sufficiently large to account for any existing and hypothetical axial flaws.

Also, based on the fabrication ultrasonic testing (UT) results, the fl aw aspect ratios that are observed are bounded by the analyzed aspect ratio of 2 for axial flaws. For exampl e, based on the avail able axial flaw UT results for the APl000 Vogtle and V. C. Swnmer units, [

t c.e This particular detected aspect ratio is bounded by the axial flaw aspect ratio of 2 analyzed in Reference 1.

Trademark Note :

APIOOO is a trademark or registered trademark of Westinghouse Electric Company LLC. its affiliates and/or its subsidiaries in the United States of America and may be registered in otl1er countries througho ut the world. All rights reserved. Unautl1orized use is strictly prohibited. Other names may be trademarks of tl1eir respective owners.

SVP_SV0_004649 ND-22-0817 Enclosure 1- VEGP 3&4-ISil -ALT- 17 Page 8 of 10 Page 36 of 38 Westinghouse Non-Propri etary Class 3 Page 3 of5 LTR-PAFM-17-6 Rev. 0 NP-Attachment January 20, 2017 Circumferential Flaw For the postulated circumferential flaw, the analysis in Reference 1 considers AR = 6 (flaw length/flaw depth). This flaw shape is a typical aspect ratio for various applications in fracture mechanics. For instance, the Pressure Temperature (P-T) limits evaluation in ASME Section XI Appendi x G also considers postulated fl aw shapes to have an aspect ratio of 6: 1. Industry experiences have also shown that the flaws found in-service are typically below AR = 6 (on the order of AR = 2 to 4 or even less) . It should be noted that the APl000 steam generator to pump DM weld region does not experience any high thermal stratification, as evident by the minimal fatigue usage discussed in Section 3 of Reference I; therefore, there is low susceptibility for any fatigue crack initiations or propagation of existing fabrication indications. Therefore, the aspect ratio of 6 is sufficient to account for any existing and hypothetical circumfe rential flaws.

Also, based on the fabrication ultrasonic testing results, the circumferential fl aw aspect ratios that are observed are bounded by the analyzed aspect ratio of 6 for circumfe rential flaws. For example, based on the available circumfe rential fl aw UT results for the APl000 Vogtle and V. C. Summer units, [

] a.c.c This particular detected aspect ratio is bounded by the circumferential aspect ratio of 6 in the analysis (Refe rence I).

OUTERDIAMETER I

'/, In.*

/

n--r-.---- -

11 11 11 I Reactor Coolant Pu mp Casing l I I Steam Generator Nozzle IC.st Austenitic Stainlm Steel j I I I {Low Alloy Steel Oad with Austenitlc Stainless Steelj Cl U I O

. :i **;1********::

1/Jt : I ' ,-~

- ~ ~)~ ; - - - - - - - -- - - - - - -- - - - - - - - - -- - - Ji Examination Volume Cross-Section INN ERDIAMETER Figure 1: Schematic of AP 1000 Steam Generator and Reactor Coolant Pump Inspection Region

SVP _SV0_ 004649 ND-22-0817 Enclosure l - VEGP 3&4-ISII -ALT-17 Page 9 of 10 Page 37 of 38 Westinghouse Non -Proprietary Class 3 Page 4 of5 LTR-PAFM- 17-6 Rev. 0 NP-Attachment January 20, 2017

2. Appendix C, Paragraph C-5300, requires that the allowable flaw depth and length be evaluated to determine acceptability. However, only flaw depth was evaluated. Additional infonnation justifying this approach is provided below.

For the Appendix C-5000 evaluation, the evaluation is per full y-plastic fracture mechanics using limit load. The limiting allowable flaw parameter for failure of this type is the flaw depth which was reported in the analysis (Reference 1). The allowable flaw length was not reported as it is not the limiting flaw parameter. However, based on the maximum end-of-evaluation allowable flaw sizes that were calculated (see Table 1), the allowable axial and circumferential flaw lengths can be calculated by multiplying the allowable flaw depths by the aspect ratios (see Table 1).

Ta bl e 1 Max1mum. Edn -ofE- va1uat1on All owable Fl aw S.1ze, Depth , an d Len~ th AR (flaw Maximum End of All owable Allowable Flaw Thickness length /flaw Evaluation Allowable Flaw Flaw Depth Flaw Length Orientation (in.)

depth) size (alt) (in.) (in.)

Axial 2 0. 75 (Reference 1, Table 3) [ ] a.c.e

[ r c,e [ ]a.c.e

]a.c.c r c.e [ ]a.c.e Circumferential 6 0.47 (Reference 1, Table 3) [ [

a = flaw depth , t = wall thickness The axial and circumferential maximum end-of-evaluation allowable flaw lengths are [

f"c.e Therefore, the detected flaw lengths are below the calculated maximum end-of-evaluation allowable flaw lengths.

If fatigue crack growth is considered, then the maximum all owable initial flaw depths and lengths for 60 years of growth are shown in Table 2.

Table 2: Maximum All owable Initial Flaw Size, Depth, and Length Accounting for 60 Years of Fatigue Crack Growth Allowable All owable Flaw Maxim um Allowable Initial Thickness AR Flaw Depth- Flaw Length-Orientation Flaw Size for 60 Years (alt) (in .)

60 years (i n.) 60 years (in.)

r,c,e ] a.c.c ] a,c,c Axial 2 0.60 (Reference 1, Fig 2) [ r [

Circumferential 6 0.30 (Reference 1, Fig 3) [ ] a.c.c [ ]'LC .C

[ r ,c.c a= flaw depth, t = wall thickness The maximum allowable initial flaw lengths for 60 years are [

svP_svo_004649 ND-22-0817 Enclosure I - YEGP 3&4-ISI I-ALT- 17 Pag e 10 of 10 Page 38 of 38 Westinghouse Non-Proprietary Class 3 Page 5 of 5 LTR-PAFM-17-6 Rev. 0 NP-Attachment January 20, 2017

Reference:

I . LTR-PAFM-16-59-P, Revision 1, "NRC RAJ Response Regarding Inspection of APl000 Vogtle Units 3 & 4 and V. C. Summer Units 2 & 3 Steam Generator to Reactor Coolant Pump Suction Nozzle Weld," November 2016.

Southern Nuclear Operating Company ND-22-0817 Enclosure 2 Vogtle Electric Generating Plant (VEGP) Units 3 & 4 Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1):

Alternative for Use of Code Case N-648-2 for lnservice Inspection of the Reactor Vessel Nozzle Inside Radius Sections (VEGP 3&4-ISl1-ALT-18)

(This Enclosure consists of 24 pages, including this cover page)

ND-22-0817 Enclosure 2 - VEGP 3&4-ISil-ALT-18 Page 2 of24 10CFR50.55a Alternative VEGP3&4-1S11-AL T-18 Proposed Alternative in Accordance with 10CFR50.55a(z)(l)

--Alternative Provides Acceptable Level of Quality and Safety-Revision 0 (Page 1 of 6)

1. ASME Code Component(s) Affected Code Class: I

Reference:

IWB-2500, Table IWB-2500- I Regulatory Guide I. I 4 7, Revision 20 Examination Category: B-D Item Number: B3 .100

Description:

Alternative for Use of Code Case N-648-2 for Inservice Inspection of the Reactor Vessel Nozzle Inside Radius Sections Component Number: Unit 3 Unit4 SV3-RPV-24A-10 I-IRS SV4-RPV-24A- I0I-IRS SV3-RPV-24B-10 I-IRS SV4-RPV-24B-I0I -IRS SV3-RPV-24C-I 0 I-IRS SV4-RPV-24C- I0I-IRS SV3-RPV-24D-101-IRS SV4-RPV-24D-10 I-IRS SV3-RPV-25A-I02-IRS SV4-RPV-25A-102-IRS SV3-RPV-25B-102-IRS SV4-RPV-25B-102-IRS SV3-RPV-26A-103-IRS SV4-RPV-26A-103-IRS SV3-RPV-26B-103 -IRS SV4-RPV-26B-103-IRS

2. Applicable Code Edition The First Interval of the Vogtle Electric Generating Plant (VEGP), Units 3 and 4 Inservice Inspection (ISI) Program is based on the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (BPV) Code,Section XI, 2017 Edition.

3. Applicable Code Requirement

Subarticle IWB-2500 requires components specified in Table IWB-2500-1 to be examined. Table IWB-2500-1 requires a volumetric examination of all Reactor Vessel nozzle inside radius sections (IRS) each inspection interval (Examination Category B-D, Item Number B3. 100). The examination volume is shown in Figure IWB-2500-7.

ASME Code Case N-648-2 (N-648-2), "Alternative Requirements for Inner Radius Examinations of Class I Reactor Vessel Nozzles,Section XI, Division I," (conditionally approved for use under Regulatory Guide 1. I 4 7, "Inservice Inspection Code Case Acceptability, ASME Section XI, Division I," Revision 20) provides an alternative to the ASME Section XI requirements stated above by allowing a VT- I visual examination in lieu of the required volumetric examination.

Regulatory Guide 1.147, Revision 20, approved N-648-2, with the following condition:

ND-22-0817 Enclosure 2 - VEGP 3&4-ISil-ALT-18 Page 3 of24 10CFR50.55a Alternative VEGP3&4-1S11-AL T-18 Proposed Alternative in Accordance with 10CFR50.55a(z)(l)

--Alternative Provides Acceptable Level of Quality and Safety--

Revision 0 (Page 2 of 6)

"This Code Case shall not be used to eliminate the preservice or inservice volumetric examination of plants with a combined operating license under 10CFR Part 52, or a plant that receives its operating license after October 22, 2015."

4. Reason for Request

In accordance with 10CFR50.55a(z)(l), relief is requested on the basis that the proposed alternative will provide an acceptable level of quality and safety.

Regulatory Guide 1.147, Revision 20, which conditionally approves Code Case N-648 -2, does not allow VEGP to utilize the code case in order to perform a VT-1 visual examination in lieu of a UT examination as currently required in Table IWB-2500-1, Examination Category B-D, Item B3 .100 during the Inservice Interval. It is the intention of Southern Nuclear Operating Company, Inc. (SNC) to adopt this Code Case for the current ISI Interval. The proposed alternative is requested to align the First ISI Interval with the performed Preservice Inspections and with the subsequent planned Inservice Inspections. This code case is used extensively in operating units.

5. Proposed Alternative and Basis for Use

SNC proposes to perform a VT-1 visual examination of the reactor vessel nozzle inside radius sections for the two outlet nozzles, the four inlet nozzles and the two direct vessel injection (DVI) nozzles using a remote underwater visual examination process that will be comparable to the historical preservice inspections. This visual examination will be conducted in accordance with the ASME BPV Code,Section XI, 2017 Edition. All of the requirements defined in Section 2, "Inservice Examinations," of Code Case N-648-2 will be applied.

The PSI Alternative VEGP3&4-PSI-ALT-07 was approved using Code Case N-648-1.

Code Case N-648-2 has since superseded Code Case N-648-1 and is conditionally approved for use under Regulatory Guide 1.147. Code Case N-648-2 was revised to include use of the code case for preservice examinations. The PSI alternative addressed the NRC condition with N-648-1 by utilizing the ASME Section XI Table IWB-3512-1 acceptance criteria, and VEGP went beyond what would currently be required of Code Case N-648-2, in that a manual UT was performed as part of the preservice examination.

Another concern with Code Case N-648-2, was that the staff requested a plant specific flaw tolerance be performed for the APlO00 nozzle at the inside radius comer. The PSI and this current ISi alternative satisfies that request by providing the flaw tolerance evaluation for the APl000 nozzles. Therefore, Code Case N-648-2 is applicable for use and PSI examinations have been performed over and above what is currently requested under N-648-2, which provides more assurance to the integrity of the nozzles.

The proposed remote VT-1 visual examination will be performed in accordance with ASME Code,Section XI, which requires that a visual examination performed instead of

ND-22-0817 Enclosure 2 - VEGP 3&4-ISil-ALT-18 Page 4 of24 10CFR50.55a Alternative VEGP3&4-1S11-AL T-18 Proposed Alternative in Accordance with 10CFR50.55a(z)(l)

--Alternative Provides Acceptable Level of Quality and Safety-Revision 0 (Page 3 of 6) an ultrasonic examination has a magnification that has a resolution sensitivity to resolve 0.044 inch (1.1 mm) lower case characters without an ascender or descender (e.g., a, e, n, v).

This technical basis addresses a VT-1 visual examination approach that includes a deterministic fracture assessment similar to that performed as a basis for Code Case N-648-2. The code alternative provides an acceptable level of quality and safety in accordance with 10CFR50.55a(z)(l) because the VT-1 visual examinations are sufficient to detect service-induced flaw mechanisms (fatigue) occurring at the inner diameter (ID) surface well before the nozzle experiences degradation of its structural integrity. The proposed remote VT-1 visual examination will cover essentially 100 percent of the ASME Code,Section XI, Figure IWB-2500-?(b) section M-N surface area for the inlet, outlet, and DVI nozzles.

Fracture Assessment A fracture assessment was performed to determine the maximum initial flaw size that will not grow beyond the allowable end of evaluation period flaw size for the life of the plant (60 years) considering Level A/B/Test conditions which were limiting in comparison to the Level CID/Test. The allowable end of evaluation period flaw sizes (depths) for the AP 1000 inlet, outlet, and DVI nozzles were determined using both linear elastic fracture mechanics (LEFM) and elastic plastic fracture mechanics (EPFM) methods.

For the most limiting case, the DVI nozzle using the LEFM analysis, the acceptance criteria for the VT-1 visual examinations using the requirements of ASME Section XI are much more limiting than the governing initial flaw depth for each nozzle and would not allow a flaw length that results in an unacceptable flaw depth during the examination period without performance of repair/replacement and regulatory review in accordance with ASME Section XI, IWB-3113 . For example, for the limiting DVI Nozzle Case, the initial limiting flaw size that would grow to the limiting flaw depth of 0.358" over a 10-year period is 0.351". Using a0.5 flaw depth to length ratio, this corresponds to a 1.14" flaw length on the surface of the cladding ((0.351" + 0.22") / 0.5), given a 0.22" cladding thickness. The proposed ASME Section XI Table IWB-3512-1 acceptance criteria of 0.144" for the maximum allowable flaw length detected during the VT-1 visual examination is much more stringent than supported by the fracture mechanics analysis of 1.14" in length.

The VT-1 visual examination acceptance criteria are conservative for the limiting LEFM case; however, it is important to note that more realistic EPFM results show that a flaw over 3" in depth for each reactor vessel nozzle can be tolerated for a 60-year plant life.

Examinations performed during fabrication of the nozzle forgings include magnetic particle examination in accordance with ASME Section III and ultrasonic examination in

ND-22-0817 Enclosure 2 - VEGP 3&4-ISil-ALT-18 Page 5 of24 10CFR50.55a Alternative VEGP3&4-1S11-AL T-18 Proposed Alternative in Accordance with 10CFR50.55a(z)(l)

--Alternative Provides Acceptable Level of Quality and Safety--

Revision 0 (Page 4 of 6) accordance with ASME Section V, Article 7 and Article 5. These examinations ensure that examination surfaces have been appropriately prepared for the application of future volumetric examinations. Following deposition of the cladding, a PT examination was performed using the acceptance standards of ASME Code Section III, NB-5350.

Following intermediate heat treatment, a UT in accordance with the examination procedure requirements of ASME Section V, using the acceptance standards of Section XI, IWB-3512 was performed with no recordable indications. Following the hydrostatic test, a PT examination was performed using the acceptance standards of ASME Code Section III, NB-5350, with no relevant indications (i.e., an indication greater than 1/16" long). These examinations, in addition to the proposed visual examination, provide assurance that existing flaws are limited in length and flaws do not exist on the cladding inner diameter surface prior to implementing a VT-1 visual examination. These examinations provide the bases for using the postulated flaw size of 0.16" (this depth correlates to a 0.32" flaw length) in the ASME Code Section III, Appendix G analysis.

Each reactor vessel satisfies ASME Section III, Appendix G requirements.

Comparison to Operating Fleet The purpose of the examination of nozzle inner radii is to detect fatigue cracking due to operation and service conditions of the component. The absence of fatigue cracking in nozzle inside radius sections during the operating history of Pressurized Water Reactor (PWR) commercial nuclear power plants, which have been inspected either ultrasonically or visually, indicates that these areas are not readily susceptible to fatigue cracking. The ability to visually detect fatigue cracks has been demonstrated successfully with probability of detection of 80% or greater.

This data is supported most recently via the joint round robin conducted by the industry Electric Power Research Institute (EPRI) and the research arm of the NRC, Pacific Northwest National Laboratory (PNNL), in a 3-phase joint project. The types of cracking the round robin was attempting to detect were much more challenging than fatigue cracking (Intergranular Stress Corrosion Cracking (IGSCC) & Irradiation Assisted Stress Corrosion Cracking (IASCC)) due to the inherent morphology of fatigue cracks versus IGSCC or IASCC. The round robin and previous operating experience clearly demonstrate that detection of cracking is primarily dependent upon the crack opening which would be greater for fatigue cracks.

Visual examination for critical reactor vessel components is routinely conducted industry wide via the Boiling Water Reactor Vessel Internals Program (BWRVIP) for BWRs and Materials Reliability Program (MRP) for PWRs. The VT-1 visual examination method as well as EVT-1 enhanced visual examination and VT-3 visual examination are applied for these critical reactor components. Based on the above, VT-1 visual examinations of the nozzle inner radii are adequate for detection of fatigue cracks should they be present.

ND-22-0817 Enclosure 2 - VEGP 3&4-ISil -ALT-18 Page 6 of24 10CFR50.55a Alternative VEGP3&4-1S11-AL T-18 Proposed Alternative in Accordance with 10CFR50.55a(z)(l)

--Alternative Provides Acceptable Level of Quality and Safety-Revision 0 (Page 5 of 6)

In addition to the above discussion, it is noted that the nozzles to which the alternative applies are fabricated from nozzle forgings, in the same manner in which the nozzles now in service in operating plants were fabricated. The material properties (yield strength, ultimate strength, and fracture toughness) for the AP 1000 nozzles are the same or better as compared to the operating fleet. The geometries are also similar, in that there are no welds in the region of the nozzle comer. Furthermore, the stresses at the nozzle comer region for the AP 1000 are similar to the operating fleet; nevertheless, a plant specific stress analysis evaluation was performed for the APlO00 nozzle comer regions based on finite element analysis. The stresses were then used to perform a plant specific flaw tolerance evaluation based on ASME Section XI (as described in the alternative request) and design basis ASME Section III Appendix G evaluation to demonstrate the structural integrity of the nozzle comer with presence of a large postulated flaw.

The water chemistry and PWR environment for the APl000 are also similar to the operating fleet. The APl 000 chemistry requirements follow the latest provided EPRI water chemistry requirements, and over time the requirements have become stricter due to the advances in instrumentation and their sensitivities. Nevertheless, the APl000 water chemistry ranges (such as pH, boron concentration, conductivity, dissolved hydrogen and oxygen) are similar to that of the operating fleet. In general, lack of oxygen in the water chemistry precludes general corrosion and wastage in the carbon steel during normal operating conditions (where primary water chemistry is controlled and which generally represents about 90% of the plant lifetime). During shutdown conditions, any potential for corrosion is prevented with the presence of stainless-steel cladding which is layered over the carbon steel base material.

During PSI, in addition to these VT-1 visual examinations, a liquid penetrant (PT) surface examination was performed at the plant site on the nozzle inside radius sections of the two outlet nozzles, the four inlet nozzles and the two DVI nozzles. The PT surface examination results were evaluated in accordance with ASME Code Section III, NB-5350. The PT examinations were performed prior to the VT-1 visual examinations.

The technical basis for this proposed alternative is included in Enclosure 1. Enclosure 1 was submitted as part of the approval process of the PSI VEGP3&4-PSI-ALT-07 Alternative, under ADAMS Accession Number MLl 7192A125, and the justification of this alternative has been proven for the life of the plants. This proposed alternative provides an acceptable level of quality and safety in accordance with 10 CFR 50.55a(z)(l).

6. Duration of Proposed Alternative

The proposed alternative is requested for the First ISi Interval for Vogtle Electric Generating Plant (VEGP), Units 3 and 4.

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--Alternative Provides Acceptable Level of Quality and Safety-Revision 0 (Page 6 of 6)

7. Precedents

Vogtle Electric Generating Plant, Units 3 and 4 Preservice Inspection Interval Proposed Alternative VEGP3&4-PSI-ALT-07 was authorized by an NRC SE dated September 25, 2018 (i.e., NRC Accession Nos. ML18263A215 and ML18263A219).

8. References

1. ASME Boiler and Pressure Vessel Code, Code Case N-648-2, "Alternative Requirements for Inner Radius Examinations of Class 1 Reactor Vessel Nozzles,Section XI, Division 1," dated September 4, 2014.

2. NRC Regulatory Guide 1.147, "Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1," Revision 20, dated December 2021.

3. ASME Boiler and Pressure Vessel Code,Section XI, Division 1, "Inservice Inspection of Nuclear Power Plant Components," 2017 Edition.

4. ASME Boiler and Pressure Vessel Code,Section III, Division 1, "Rules for Construction of Nuclear Facility Components," 1998 Edition through the 2000 Addenda.

9. Enclosure Enclosure 1: ND-17-1121, Technical Basis for the Alternative Request on Inservice Inspection Requirements for Reactor Vessel Nozzle Inner Radius Sections (Previously Enclosure 2 under Alternative VEGP3&4-PSI-ALT-07)

ND-22-0817 Enclosure 2 - VEGP 3&4-ISil -ALT-18 Page 8 of24 10CFR50.55a Alternative VEGP3&4-ISI1-ALT-18 Proposed Alternative in Accordance with 10CFR50.55a(z)(l)

-Alternative Provides Acceptable Level of Quality and Safety--

Revision 0 Enclosure 1 ND-17-1121, Technical Basis for the Alternative Request on Preservice Inspection Requirements for Reactor Vessel Nozzle Inner Radius Sections (Previously Enclosure 2 under Alternative VEGP3&4-PSI-ALT-07)

ND-22-0817 Enclosure 2 - VEGP 3&4-ISII-ALT- 18 Page 9 of24 ND-17-1121 Technical Basis for the Alternative Request on Preservice Inspection Requirements for Reactor Vessel Nozzle Inner Radius Sections TECHNICAL BASIS FOR THE ALTERNATIVE REQUEST ON PRESERVICE INSPECTION REQUIREMENTS FOR REACTOR VESSEL NOZZLE INNER RADIUS SECTIONS 1.0 Introduction ASME Code Section XI Code Case (CC) N-648-1 [1] allows for the use of a VT-1 visual examination in lieu of the volumetric examination requirement defined in ASME Code Section XI, Table IWB-2500-1 , Examination Category B-D, Item No. B3.100 [2]. This code case is conditionally accepted by the NRC in Regulatory Guide 1.147 [3]. The condition is that the allowable flaw length criteria of ASME Code Section XI , Table IWB-3512-1 must be used with limiting assumptions on the flaw aspect ratio. CC N-648-1 applies only to inservice inspection (ISi) .

The technical basis for CC N-648-1 is documented in a paper prepared for and presented at the ASME 2001 Pressure Vessels and Piping Conference [4]. The key arguments to justify elimination of the volumetric ISi requirements are good inspection history, a large flaw tolerance , and a risk argument concluding that there is negligible change in core damage frequency with the elimination of the inspection . The logic for the VT-1 visual examination is that service-induced flaw mechanisms (fatigue) will be associated with the inner diameter (ID) surface of the cladding and that the VT-1 examinations are sufficient to detect such mechanisms occurring at the ID surface well before the nozzle suffers degradation of its structural integrity.

It is proposed to extend the application of VT-1 visual examination to the preservice inspection (PSI) subject to the following requirements :

The surface M-N shown in Figure IWB-2500-7 sketches (a) through (d) is examined using a surface examination method and shall meet the Section Ill fabrication acceptance standards at least once after the Construction Code hydrostatic test. The surface examination is performed prior to the preservice VT-1 visual examination .

The volume O-P-Q-R shown in Figure IWB-2500-7 sketches (a) through (d) is examined using a manual volumetric examination method and shall meet the Section XI acceptance standards at least once after the Construction Code hydrostatic test.

The appropriate surface is prepared in accordance with IWA-2200(b) for application of a future volumetric examination in accordance with Table IWB-2500-1, Examination Category B-D.

1

ND-22-0817 Enclosure 2 - VEGP 3&4-ISII-ALT-18 Page 10 of 24 ND-17-1121 Technical Basis for the Alternative Request on Preservice Inspection Requirements for Reactor Vessel Nozzle Inner Radius Sections

An evaluation that includes the following is performed :

o Review of the fabrication examination history for the nozzle inner radius region o Verification that the nozzle of interest meets the requirements of Section Ill, Nonmandatory Appendix G.

This technical basis addresses a VT-1 visual examination approach that includes a deterministic fracture assessment similar to that performed as a basis for CC N-648-1, provides a description and justification of a preservice inspection process that addresses the requirements provided above, and will provide an acceptable level of quality and safety in accordance with 10 CFR 50.55a(z)(1).

2.0 PWR Nozzle Inner Radius Section Inspection History in Industry The ASME Code Section XI 1971 Edition through the 2015 Edition requires volumetric examination of the reactor vessel inner radius section. The original requirement for an examination of this region was developed as a result of cracking in a non-nuclear vessel that occurred around the time when the ASME Code Section XI inspection requirements were being established [6].

Up until the implementation of ASME Code Section XI Code Case N-648-1 after 20011, volumetric examinations of PWR reactor vessel nozzle inner radius sections were conducted as required by ASME Section XI using the ultrasonic test method . No recordable flaw indications were detected [6]. Subsequently, enhanced VT-1 visual examinations with a resolution capability of distinguishing a 1-mil wire or crack have been applied to PWR reactor vessel nozzle inner radius sections. Again , no recordable flaw indications have been detected.

3.0 Reactor Vessel Nozzle Inner Radius Section Design and Fabrication Inspection History The reactor vessel and the reactor vessel nozzles are designed in accordance with the ASME Code Section Ill , Subsection NB [7] . The reactor vessel nozzles are fabricated of 1

Code Case N-648-1 was conditionally approved by the NRC in Regulatory Guide 1.147, Revision 13 issued in June 2003 .

2

ND-22-0817 Enclosure 2 - VEGP 3&4-ISII -ALT- 18 Page 11 of 24 ND-17-1121 Technical Basis for the Alternative Request on Preservice Inspection Requirements for Reactor Vessel Nozzle Inner Radius Sections SA-508, Grade 3, Class 1 [27] ferritic steel forgings clad on the inner diameter surface with multiple layers of stainless steel cladding (Type 309L first layer and Type 308L subsequent layers) . The AP1000 reactor vessel has two outlet nozzles, four inlet nozzles, and two direct vessel injection (DVI) nozzles. Figure 1 through Figure 3 show elevation view cross-sections of the three nozzle types , respectively.

Weld

[fen-itic J NOULEltlMER RADI USSECTION S,a;nle,, Steel O*d<lng \

Elevation Vie\v Uppe r Shell

[fenitic steel]

Figure 1: AP1000 Reactor Vessel Outlet Nozzle - Elevation View Cross-Section 3

ND-22-0817 Enclosure 2 - VEGP 3&4-ISil -ALT- 18 Page 12 of24 ND-17-1121 Technical Basis for the Alternative Request on Preservice Inspection Requirements for Reactor Vessel Nozzle Inner Radius Sections Weld

[ferrit ic )

NOZZLE IHIIER RADIUS SECTION St.ainess St eel O-idcin,g ln~*t NmU!

[fe rritica:eef)

RPV Upper Shell (ferritic stee l]

I Elevat ion View I Figure 2: AP1000 Reactor Vessel Inlet Nozzle - Elevation View Cross-Section 4

ND-22-0817 Enclosure 2 - YEGP 3&4-ISI 1-AL T- 18 Page 13 of24 ND-17-1121 Technical Basis for the Alternative Request on Preservice Inspection Requirements for Reactor Vessel Nozzle Inner Radius Sections

-*- * -*- - *- *- *-*-*-*-*- *- *- *- *- *- *- *- *~ *- *-

c==.:rr:::::::-----=======-

Eleva tion View Figure 3: AP1000 Reactor Vessel Direct Vessel Injection (DVI) Nozzle - Elevation View Cross-Section In accordance with ASME Code Section Ill, NB-2540, the nozzle forgings are subject to magnetic particle examination over all external surfaces and accessible internal surfaces, and ultrasonic examination of the nozzle volume in accordance with the ASME Code Section V, Article 7 and Article 5, respectively [8]. The ultrasonic test requirements are enhanced by the Westinghouse material specification for SA-508 forging materials. Such enhancements include the implementation of Supplementary Requirement S2 of SA-508 [9] that requires the use of a higher sensitivity straight beam examination calibrated on 1/4-inch diameter flat-bottomed holes rather than the forging back surface, recording and investigating of angle beam indications equal to or exceeding 20% of the reference level rather than equal to or exceeding 50% of the reference level, and specifically identifying all recordable angle beam indications located near a surface (within 15% of the wall thickness) and/or all indications that display crack-like characteristics for separate disposition.

In accordance with ASME Code Section Ill , NB-5120 (d) , the nozzle base metal surface is examined by the magnetic particle method prior to the deposition of the stainless steel 5

ND-22-0817 Enclosure 2 - VEGP 3&4-ISII-ALT-18 Page 14 of24 ND-17-1121 Technical Basis for the Alternative Request on Preservice Inspection Requirements for Reactor Vessel Nozzle Inner Radius Sections cladding using the acceptance standards of ASME Code Section Ill, NB-5340. After the cladding is deposited, the cladding surface is examined by liquid penetrant method using the acceptance standards of ASME Code Section Ill, NB-5350. The cladding is also subject to an ultrasonic examination for lack of bond as defined in the Westinghouse fabrication specification.

After completion of welding and the intermediate heat treatment but before the post-weld heat treatment, the ultrasonic test method is applied for the examination of the inlet, outlet and DVI nozzle inner radius section volumes as defined by ASME Code Section XI, Figure IWB-2500-?(b) . The nozzle inner radius section volumes are shown in Figures 1 through 3. These examinations are conducted from the inside and outside diameter surfaces in accordance with the examination procedure requirements of ASME Code Section V, Article 4 [8] and using the acceptance standards of ASME Code Section XI, IWB-3512. These are mandatory supplemental requirements defined in the Westinghouse fabrication specification.

After the vessel hydrostatic test, the Westinghouse fabrication specification requires a liquid penetrant examination of all internal vessel surfaces including the stainless steel cladding in the nozzle inner radius sections using the acceptance standards of ASME Code Section Ill, NB-5350. This specification also requires a repeat of the ultrasonic test method on the nozzle inner radius section volumes applied before the post-weld heat treatment including the examinations from the inner and outer diameter surfaces.

The examinations described above were applicable to the reactor vessel nozzles of Vogtle Units 3 and 4 and V.C. Summer Units 2 and 3.

For these four units, the post-hydrostatic test liquid penetrant examinations detected no relevant flaw indication of cracking or linear indication [10, 11, 12, 13]. A relevant indication is defined as being greater than 1/16-inch long.

For these four units, the post-hydrostatic test nozzle inner radius section ultrasonic examinations of the two outlet nozzles, four inlet nozzles and two DVI nozzles of each unit detected no recordable indications [14, 15, 16, 17]. The inner and outer diameter surface applied ultrasonic examinations consisted of the techniques defined in Table 1.

6

ND-22-081 7 Enclosure 2 - VEGP 3&4-ISII -ALT-18 Page 15 of24 ND-17-1121 Technical Basis for the Alternative Request on Preservice Inspection Requirements for Reactor Vessel Nozzle Inner Radius Sections Table 1: Ultrasonic Test Techniques Applied in the Shop on the Inlet, Outlet and DVI Nozzle Inner Radius Sections Prior to the Vessel Post-Weld Heat Treatment and After the Hydrostatic Test Applied Test Test Refere nce Nozzle Inner Surface Test Mode Surface Angle(s) Frequency Sensitivity Radius Section Outside r1I,r3I 27°, 30 °, Shear 2.25 MHz ID notch (2% T) Inlet and Outlet Nozzles 45 ° Wave Shear Outside PJ,[3I 13°, 22 ° 2.25 MHz ID notch (2% T) DVI Nozzle Wave Transmit-receive ID notch Inlet, Outlet and DVI Inside r41 70 ° 2 MHz longitudinal (2.5% a/t)r2I Nozzles wave Note 1: 0° transducer applied for the detection of laminar flaw indications that would limit or affect the interpretation of the angle beam examination results .

Note 2: Notch depth consistent with ASME Code Section XI , Table IWB-3512-1 for inside comer region .

Note 3: Examinations in two circumferential directions around nozzles .

Note 4 : Examinations in four directions, two axial and two circumferential directions .

4.0 Section Ill, Appendix G Verification The AP1000 reactor vessel was evaluated for its ability to protect against non-ductile failure in accordance with ASME Code Section Ill , Appendix G [18] requirements for postulated defects. The inlet nozzle, outlet nozzle and DVI nozzle regions were part of this linear elastic fracture mechanics (LEFM) evaluation . The fracture mechanics evaluation considered the Level A/B service condition , Level C/D service condition and Test Condition (at 70°F and at 110°F) design transients and mechanical loads.

The results demonstrate that the maximum K1 values , resulting from the design transients and mechanical loads, meet the requirements of ASME Code Section Ill, Appendix G for the postulated flaw sizes. The AP1000 reactor vessel is in compliance with ASME Code Section Ill , Appendix G. To meet these requirements, flaw sizes smaller than one-quarter of the section thickness were assumed . For the reactor vessel nozzle inner radius regions, the smallest postulated flaw size was 0.16-inch at a hydrostatic test temperature of 70°F. Such a small postulated flaw was justified based on the manufacturing inspections described in Section 3.0 and the ultrasonic and visual examinations to be performed prior to service as described in Section 6.0.

5.0 Fracture Assessment Reference [4] provides a basis to eliminate inservice volumetric examinations at the inner radius of reactor vessel nozzles for the operating reactor vessels in the US. The American Society of Mechanical Engineers (ASME) approved Code Case N-648-1 [1]

7

ND-22-0817 Enclosure 2 - VEGP 3&4-ISII -AL T- 18 Page 16 of24 ND-17-1121 Technical Basis for the Alternative Request on Preservice Inspection Requirements for Reactor Vessel Nozzle Inner Radius Sections based on the results documented in [4]. At the time of publication of [1] , the Code Case was only applicable to operating plants in the US. The fracture assessment results documented in this section support the technical basis for the AP1000 plant design.

This includes calculation of the end of evaluation period flaw sizes for the AP1000 inlet, outlet and direct vessel injection (DVI) nozzles as well as fatigue crack growth analyses using the rules of ASME Section XI [2].

The allowable end of evaluation period flaw sizes (depths) for the AP1000 inlet, outlet, and DVI nozzles were determined using both linear elastic fracture mechanics (LEFM) and elastic plastic fracture mechanics (EPFM) methods . The LEFM flaw tolerance calculations were performed per ASME Section XI IWB-3600 and Appendix A, and the EPFM method followed the guidelines of Code Case N-749 [5]. Fatigue crack growth (FCG) analyses were also performed in order to determine the maximum initial flaw size that will not grow beyond the allowable end of evaluation period flaw size within the life of the plant (60 years) considering Level A/B/Test conditions . In addition, FCG analyses were also performed to determine the maximum initial flaw size for a 10 year period using LEFM only. In all cases the crack growth law for ferritic steels not susceptible to environmentally assisted cracking (EAC) given in Code Case N-643-2 [26] was used for the FCG calculations.

Table 2 shows the fracture assessment results for Level A/B/Test conditions for all three nozzle types using both LEFM and EPFM methods. The LEFM method is very conservative because it does not take into account the ductile behavior of the nozzle material , due to the lack of constraint present in this geometry. The EPFM results listed in Table 2 were produced using Code Case N-749 [5] and provide a more realistic fracture assessment considering the resistance to crack extension of the ductile nozzle material.

For the LEFM results , the DVI nozzle design produced the smallest end of evaluation period flaw size (0.358 inch) , as well as the most limiting FCG result (0.326 inch) for a 60 year operating life. The results for ten years of operation show tolerance for slightly larger flaws and demonstrate the flaw sizes that might be of concern between inspection intervals based on conservative LEFM evaluations.

The EPFM evaluations demonstrate tolerance for much larger flaws , as shown in Table 2. The most limiting end of evaluation period flaw size is 4.5 inches for the DVI nozzle. However, the most limiting FCG result occurs for the outlet nozzle with a flaw depth of 3.088 inches. In all cases, tolerance for flaws over three inches in depth is demonstrated for 60 years of operation . Because the 60 year results demonstrate tolerance for such large flaws , it was not necessary to evaluate a 10 year period as was done for the LEFM cases .

The end of evaluation period flaw sizes for Level CID conditions were also determined for each nozzle type using the LEFM method. As can be seen in Table 3, the Level C/D 8

ND-22-0817 Enclosure 2 - VEGP 3&4-ISI l-ALT-18 Page 17 of 24 ND-17-1121 Technical Basis for the Alternative Request on Preservice Inspection Requirements for Reactor Vessel Nozzle Inner Radius Sections flaw evaluation results are not limiting in comparison to the Level A/B!Test LEFM results reported in Table 2. For all cases listed in Table 3, the limiting flaw sizes are over 3 inches.

These initial flaw size results for 60 years are considered to be acceptable based on Section Ill flaw acceptance criteria prior to the components being placed into service.

The largest permissible flaw length for magnetic partial examination per NB-2545 for forgings is 3/16 inch . The analyses were performed using the nozzle corner, quarter-circular stress intensity factor solution from API 579-1 [25] with the built-in assumed length-to-depth ratio of 2. Thus , the depth corresponding to a 3/16 (0.1875) inch flaw length would be 3/32 (0.094) inch . Additionally, the in-process and post-hydrostatic test UT examinations of the nozzle inner radius sections from the ID and OD surfaces (described in Section 3.0) detected no indications that may have appeared after cladding of the ID surface. Therefore, any flaw that would have been placed into service would have a depth less than the limiting flaw size reported in Table 2, even for the conservative LEFM cases .

9

ND-22-0817 Enclosure 2 - VEGP 3&4-ISII-ALT-18 Page 18 of24 ND-17-1121 Technical Basis for the Alternative Request on Preservice Inspection Requirements for Reactor Vessel Nozzle Inner Radius Sections Table 2: End of Evaluation Period Flaw Size and Fatigue Crack Growth Results for Inlet, Outlet and DVI Nozzles (Level A/8/Test Conditions)

End of FCG Results FCG Results Evaluation Period= 10 Period= 60 LEFM/EPFM Component\Location Period Flaw Size Years Years (in) (in) (in)

Inlet Nozzle\Cut 5 1.034 0.946 0.663 Inlet Nozzle\Cut 6 0.988 0.944 0.786 LEFM Outlet Nozzle\Cut 5 1.151 1.066 0.793 Outlet Nozzle\Cut 6 0.922 0.884 0.766 DVI Nozzle\Cut 8 0.362 0.356 0.335 DVI Nozzle\Cut 9 0.358 0.351 0.326 Inlet Nozzle\Cut 5 7.0 NIA 4.542 Inlet Nozzle\Cut 6 5.0 NIA 4.195 EPFM Outlet Nozzle\Cut 5 6.0 NIA 3.595 Outlet Nozzle\Cut 6 4.0 NIA 3.088 DVI Nozzle\Cut 8 5.0 NIA 4.652 DVI Nozzle\Cut 9 4.5 NIA 4.132 Table 3: End of Evaluation Period Flaw Size for Level C/O Conditions Using LEFM Method End of Evaluation Nozzle/Cut Period Flaw Size (in.)

lnlet/5 7.200 lnlet/6 7.200 Outlet/5 9.480 Outlet/6 9.480 DVl/8 5.260 DVl/9 3.477 6.0 Preservice Inspection Process for Reactor Vessel Nozzle Inner Radius Sections The ASME Code Section XI preservice inspection (PSI) of the nozzle inner radius sections for the two outlet nozzles, the four inlet nozzles and the two DVI nozzles will be done using a VT-1 visual examination method using an underwater camera system attached to a submersible. This process will be comparable to the subsequent inservice inspections. This visual examination will be conducted in accordance with the ASME 10

ND-22-0817 Enclosure 2 - VEGP 3&4-ISII-ALT-18 Page 19 of 24 ND-17-1121 Technical Basis for the Alternative Request on Preservice Inspection Requirements for Reactor Vessel Nozzle Inner Radius Sections Code Section XI , 2007 Edition with the 2008 Addenda. The allowable flaw length criteria of ASME Code Section XI , Table IWB-3512-1 with a flaw aspect ratio (al/) of 0.5 will be applied for any detected flaw indication . This exception is consistent with the condition defined in NRC Regulatory Guide 1.147 for Code Case N-648-1 . Table 4 provides the acceptance standards for the VT-1 visual examination specific to the AP1000 reactor vessel nozzles.

However, prior to the VT-1 visual examination PSI , liquid penetrant (PT) surface examinations will be performed at the plant site on the nozzle inner radius sections of the two outlet nozzles , the four inlet nozzles and the two DVI nozzles . The liquid penetrant examinations will be conducted in accordance with ASME Code Section XI, IWA-2222 using the ASME Code Section Ill , NB-5350 acceptance standards. This is a repeat of the surface examinations performed in the manufacturer's shop as described in Section 3.0 after the Construction Code hydrostatic test. These repeat surface examinations are applied to ensure that no relevant surface-breaking flaws are present on the cladding surfaces prior to service. The PT examination report is to be included in the preservice inspection (PSI) documentation package.

After the PT examinations, manual ultrasonic examinations (UT) will be conducted at the plant site. These UT examinations will be applied from the inner diameter surface using two opposing circumferential beam directions around the nozzle inner radius sections of the two outlet nozzles, the four inlet nozzles and the two DVI nozzles. Dual focused 70-degree transmit-receive longitudinal wave transducers with acoustic focusing at or near the clad/base metal interface will be used to interrogate the nozzle inner radius section examination volume as defined in ASME Code Section XI , Figure IWB-2500-7(b) for radial-axial flaws (see Figures 1 through 3). The ultrasonic examination procedure requirements will be in accordance with ASME Code Section XI , Appendix Ill as supplemented by ASME Code Section XI , Appendix I Supplements 1 - 8, 1O and 11 .

These supplements are:

Supplement 1 - Calibration Block Material and Thickness

Supplement 2 - Calibration Blocks for Clad Welds or Components

Supplement 3 - Calibration Blocks for Examination of Parts with Curved Surfaces

Supplement 4 - Alternative Weld Calibration Block Design

Supplement 5 - Electronic Simulators

Supplement 6 - Pulse Repetition Rate

Supplement 7 - Instrument Calibration

Supplement 8 - Scan Overlap and Search Unit Oscillation

Supplement 10 - Recording Criteria

Supplement 11 - Geometric Indications 11

ND-22-0817 Enclosure 2 - VEGP 3&4-ISII-ALT-18 Page 20 of24 ND-17-1121 Technical Basis for the Alternative Request on Preservice Inspection Requirements for Reactor Vessel Nozzle Inner Radius Sections The reference sensitivity will be established on a radial -axial notch at the inside corner region with a depth equal to an 'a/f of 2.5% consistent with the ASME Code Section XI, Table IWB-3512-1 allowable planar flaw size.

It is noted that dual focused 70-degree transmit-receive longitudinal wave transducers have proven to be effective at detecting near surface flaws initiating at the cladding surface or at the clad/base metal interface [19 - 24].

These ultrasonic examinations ensure that the appropriate surfaces have been prepared for application of future volumetric examinations in accordance with ASME Code Section XI , Table IWB-2500-1 , and they provide a baseline volumetric examination of the nozzle inner radius section volumes . The UT examination report is to be included in the preservice inspection (PSI) report.

Table 4: AP1000 Reactor Vessel Nozzle Inside Corner Region VT-1 Visual Examination Acceptance Standards in Accordance with ASME Code Section XI, Table IWB-3512-1 Component Max. 'alt' Max. 'a' Nozzle Max. 'all' Max. '/' Allowed£51 Thickness111 Allowedl31 Allowed141 Description Allowed121 [inch]

[inch] [%] [inch]

tn1 = 12.03 Outlet tn2 = 7.85 0.5 2.5 0.1 96 0.392 ts = 10.15 tn1 = 11 .91 Inlet tn2 =7.68 0.5 2.5 0.192 0.384 ts =10.15 tn1 = 2.87 DVI =

tn2 2.87 0.5 2.5 0.072 0.144 ts= 10.15 Note 1: Thickness is the smallest of the three thicknesses shown in ASME Code Section XI, Figures IWB-2500-7(a) and

-7(b) as defined by Table IWB-3512-2. The smallest component thickness is shown in bold print.

Note 2: Based on Reg . Guide 1.147, Rev. 17 condition to Code Case N-648-1 and approach defined in this technical basis ; 'all' is the flaw depth / flaw length ratio.

Note 3: Based on ASME Code Section XI, Table IWB-3512-1 for Inside Corner Region ; applicable for nominal wall thickness ranging from 2.5-inches or less to 12-inches; 'a/f is the flaw depth / component thickness ratio . This table was defined in the Reg , Guide 1.147 condition on CC N-648-1 and is the same approach defined in this technical basis .

Note 4 : Calculated as 2.5% of the smallest component thickness or 0.025 times the smallest component thickness.

Note 5: Calculated as 'a'/0.5 .

7.0 Conclusions The following conclusions can be made:

1. The combination of the post-hydrostatic test surface (PT) examinations performed at the manufacturer's facility and at the plant site ensures that fabrication flaws, particularly cracks and linear flaws , do not exist on the cladding inner diameter 12

ND-22-0817 Enclosure 2- VEGP 3&4-ISII-ALT-18 Page 21 of24 ND-17-1121 Technical Basis for the Alternative Request on Preservice Inspection Requirements for Reactor Vessel Nozzle Inner Radius Sections surface prior to implementing the VT-1 visual examination method. The ASME Code Section Ill , NB-5340 acceptance standards consider all cracks and linear flaws greater than 1/16-inch long to be unacceptable.

2. The combination of the post-hydrostatic test volumetric (UT) examinations performed at the manufacturer's facility and at the plant site ensures that the appropriate surfaces have been prepared in accordance with ASME Code Section XI ,

IWA-2200(b) for application of future volumetric examinations. This is consistent with the Owner's responsibil ity to provide adequate design and access provisions for periodic inservice inspection in compliance with ASME Code Section Ill ,

NCA-3220(r), ASME Code Section XI , IWA-1400(b) and IWA-1500, and 10 CFR 50.55a(g)(3)(i). The UT examinations performed at the plant site also provide for a baseline volumetric examination of the nozzle inner radius section volumes .

3. The fabrication examination history for the nozzle inner radius sections have been reviewed and documented. This examination history includes the required ASME Code Section Ill, NB-2500 and NB-5000 examinations of the forging material and cladding as supplemented by Westinghouse requirements for UT of the cladding for lack of bond, in-process and post-hydrostatic test UT examinations applied from the ID and OD surfaces for flaws in the base metal, and post-hydrostatic test PT examinations of the cladding surfaces for flaws on the ID surface. Such examinations support the maximum postulated flaw sizes used in the ASME Code Section Ill, Appendix G evaluation.

4. The nozzles meet the requirements of ASME Code Section Ill , Appendix G.

5. A deterministic fracture mechanics assessment has shown that the governing initial flaw size for the AP1000 inlet, outlet and DVI nozzles over a 10 year period consistent with the 10-year inspection interval is 0.351-inch deep within the underlying nozzle base metal , using very conservative LEFM methods. The limiting flaw size for a 60 year period is 0.326 inch, also using LEFM . More realistic EPFM results show that a flaw over 3 inches in depth can be tolerated for a 60 year plant life. The flaw depth to length aspect ratio consistent with the condition on Code Case N-648-1 in Regulatory Guide 1.147 is a/I= 0.5. Thus for a flaw depth (a) of 0.326-inch, the flaw length(/) is 0.652-inch for the base metal.

6. The governing initial flaw depth of 0.326-inch in the base metal (calculated for the DVI nozzle as indicated on Table 2) corresponds to a flaw length of 1.09-inches on the surface of the cladding given the nominal clad thickness of 0.22-inch, the total flaw depth (a) of 0.546-inch (0.326 + 0.22) and a 0.5 flaw depth to flaw length (a//)

aspect ratio. Table 4 indicates that for the VT-1 visual examination of the DVI nozzle 13

ND-22-0817 Enclosure 2 - VEGP 3&4-ISI I-ALT-18 Page 22 of24 ND-17-1121 Technical Basis for the Alternative Request on Preservice Inspection Requ irements for Reactor Vessel Nozzle Inner Radius Sections the allowable flaw length is 0.144-inch which is smaller than the governing initial flaw length of 1.09-inch given a 0.5 flaw depth to flaw length aspect ratio . A VT-1 visual examination finding exceeding this allowable acceptance standard would result in repair/replacement and reexamination, and regulatory review in accordance with ASME Code Section XI, IWB-3113 and IWB-3114 to ensu re fitness for service .

8.0 References

1. ASM E Boiler and Pressure Vessel Code, Case N-648-1 : Alternative Requirements for Inner Radius Examinations of Class 1 Reactor Vessel Nozzles,Section XI ,

Division 1, approved September 7, 2001 .

2. ASME Boiler and Pressure Vessel Code,Section XI : Rules for lnservice Inspection of Nuclear Power Plant Components , 2007 Edition with the 2008 Addenda.

3. U.S. Nuclear Regulatory Commission , Regulatory Guide 1.147: lnservice Inspection Code Case Acceptability, ASME Section XI , Division 1, Revision 17.

4. W . H. Bamford et al , 'Technical Basis for Elimination of Reactor Vessel Nozzle Inner Radius Inspections,' Proceedings of ASME 2001 Pressure Vessels and Piping Conference , Atlanta, GA.

5. ASME Boiler and Pressure Vessel Code, Case N-749: Alternative Acceptance Criteria for Flaws in Ferritic Steel Components Operating in the Upper Shelf Temperature Range Section XI , Division 1, March 16, 2012 .

6. W .H. Bamford, D. Kurek and K. Jacobs, Westinghouse WCAP-15262 : Technical Basis for Elimination of Reactor Vessel Nozzle Inner Radius Inspections, Indian Point Unit 3, July 1999 (Accession Number: ML100361141).

7. ASME Boiler and Pressure Vessel Code, Section Ill , Division 1 - Subsection NB Class 1 Components : Rules for Construction of Nuclear Power Plant Components ,

1998 Edition through 2000 Addenda .

8. ASME Boiler and Pressure Vessel Code, Section V: Nondestructive Examination ,

1998 Edition through 2000 Addenda .

9. ASME/ASTM SA-508: Specification for Quenched and Tempered Vacuum-Treated Carbon and Alloy Steel Forgings for Pressure Vessels (including S2: Ultrasonically Testing - Reference Block Calibration (for examining sections 24 in. [610mm] thick or less)) .

14

ND-22-0817 Enclosure 2 - VEGP 3&4-ISil -ALT-18 Page 23 of24 ND-17-1121 Technical Basis for the Alternative Request on Preservice Inspection Requirements for Reactor Vessel Nozzle Inner Radius Sections

10. Doosan Heavy Industries & Construction, P120706-014-001 : PT Report for the Examination of All Inside Surfaces of the Reactor Vessel (After Hydrostatic Test) for Vogtle Unit 3, July 2012.

11. Doosan Heavy Industries & Construction, P140103-031-001 : PT Report for the Examination of All Inside Surfaces of the Reactor Vessel (After Hydrostatic Test) for Vogtle Unit 4, January 2014.

12. Doosan Heavy Industries & Construction, P120905-009-001: PT Report for the Examination of All Inside Surfaces of the Reactor Vessel (After Hydrostatic Test) for V.C. Summer Unit 2, September 2012.

13. Doosan Heavy Industries & Construction, P150330-027-001: PT Report for the Examination of All Inside Surfaces of the Reactor Vessel (After Hydrostatic Test) for V .C. Summer Unit 3, April 2015.

14. Doosan Heavy Industries & Construction, U120706-021-001 : UT Report for the Examination of Vessel Girth Welds, Nozzle to Shell Welds and Nozzle Inner Radius Regions (After Hydrostatic Test) for Vogtle Unit 3, July 2012.

15. Doosan Heavy Industries & Construction, U140107-017-002: UT Report for the Examination of Nozzle Inner Radius Regions (After Hydrostatic Test) for Vogtle Unit 4, January 2014.

16. Doosan Heavy Industries & Construction, U 120907-022-002: UT Report for the Examination of Nozzle Inner Radius Regions (After Hydrostatic Test) for V.C Summer Unit 2, September 2012.

17. Doosan Heavy Industries & Construction, U150330-031-001: UT Report for the Examination of Nozzle Inner Radius Regions (After Hydrostatic Test) for V.C Summer Unit 3, April 2015.

18. ASME Boiler and Pressure Vessel Code, Section Ill, Division 1, Appendix G:

Protection Against Nonductile Failure, 1998 Edition through 2000 Addenda .

19. S.Crutzen, C. Vinche, Ph. Doubret and N. Haines, "The Evaluation of the PISC II Round Robin Tests," 8th International Conference on NOE in the Nuclear Industry, Kissimmee, Florida, November 17 - 20, 1986.

15

ND-22-0817 Enclosure 2 - VEGP 3&4-ISI I-ALT-18 Page 24 of24 ND-17-1121 Technical Basis for the Alternative Request on Preservice Inspection Requirements for Reactor Vessel Nozzle Inner Radius Sections

20. D.L. Lock, K.J . Cowburn and B. Watkins, "The Results Obtained in the UKAEA Defect Detection Trials on Test Pieces 3 and 4," Nuclear Energy: Journal of the British Nuclear Energy Society, Volume 22, Number 5, October 1983.

2 1. D.B. Langston and R. Wilson , "An Assessment of the Defect Detection Capability of the Ultrasonic Inspection Techniques Used by the CEGB in the UKAEA Defect Detection Trials ," International Journal - Pressure Vessel & Piping, Volume 23, 1986.

22. J.M. Coffey, R.K. Chapman , J.M . Wrigley and K.J . Bowker, Report No.

NWR/SSD/84/0009/E: Ultrasonic Examination of the Near Surface Regions of the Reactor Pressure Vessel , Sizewell B Power Station, January 1984.

23. T.T. Taylor, S.L. Crawford , S.R. Doctor and G.J. Posakony, NUREG/CR-2878:

Detection of Small-Sized Near-Surface Under-Clad Cracks in U.S. Reactor Pressure Vessels , January 1983.

24. P. Sermadiras and J.P. Launay , EPRI Report NP-2841 : Results of EDF/Framatome Underclad Crack Detection Methods, January 1983.

25. API 579-1 , Second Edition, "Fitness-For-Service," June 5, 2007, American Society of Mechanical Engineers.

26. ASME Boiler and Pressure Vessel Code, Case N-643-2: Fatigue Crack Growth Rate Curves for Ferritic Steels in PWR Water Environment Section XI , Division 1, May 4, 2004.

27. VEGP 3&4 UFSAR Revision 6, Table 5.2- 1 16

ND-22-0817, Enclosures 1 and 2: Proposed Alternative Inspection Requirements for Steam Generator Nozzle to Reactor Coolant Pump Casing Welds and Alternative for Use of Code Case 648-2 for Inservice Inspection

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ND-22-0817, Enclosures 1 and 2: Proposed Alternative Inspection Requirements for Steam Generator Nozzle to Reactor Coolant Pump Casing Welds and Alternative for Use of Code Case 648-2 for Inservice Inspection

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